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Highlights • Design of Toroidal Field (TF) coil structure in Alborz Tokamak. • Strength analysis of the coil structure and its support. • Report of the structure manufacturing and mounting. Abstract The structures which support the coils and other devices play an essential role in the performance of Tokamaks. Since Toroidal Field (TF) coil movement results in the magnetic field variation and plasma fluctuation, a robust structure has to be designed to prevent excessive TF coil deformations. Other issues such as the selection of a non-ferromagnetic material should be obeyed in the structure design. In this paper, the design, analysis and manufacture of the support structure for TF coils of Alborz Tokamak have been reported. The TF coil support system was designed with regard to all the technical points. After designing the structure, the strength of the system was investigated, using the Abaqus software; and to determine the input electromagnetic loads, Comsol program was used. The result of the analysis indicated the appropriate strength of the system. Accordingly, the support structures were manufactured. The primary shapes of the structures were constructed via casting and then the structure shape was finalized by Computerized Numerical Control (CNC) machining. During manufacturing, attempts were made to maximize the dimensional precision. The detailed report of the mentioned steps has been presented in this paper. [ABSTRACT FROM AUTHOR]

Highlights • This paper summarizes the development of components for ITER Edge Thomson scattering measurement system. • The beam combiner allows to minimize neutron streaming through the port plug penetrations used for laser beam injection. • A new type of beam dump to withstand harsh thermo-mechanical loads was designed and verified its concept. • A method to align the collection optics field of view was proposed and verified by ray-tracing simulation. Abstract The in-vessel components for the edge Thomson scattering measurement system in ITER (ETS) will be exposed to high thermal and electromagnetic loads, however, their maintenance will be limited because the ITER radiological environment will not allow operators to align the optics near the vacuum vessel. This paper summarizes the development of components, in particular, the beam combiner so as to minimize neutron streaming through the port plug penetrations used for las er beam injection and the strengthening of the beam dump so it can withstand harsh thermal and mechanical loads. Furthermore, a method to align the collection optics field of view was proposed and verified by ray-tracing simulation. The resolution and accuracy of which were approximately 0.5 mm and 0.1 mm, respectively. These values were much finer than the expected characteristic length of electron temperature and density gradients in the edge plasma of ITER. [ABSTRACT FROM AUTHOR]

This paper briefly presents the results of neutronics analysis performed in support for the design development of the ITER diagnostics cabling system including its cable looms. The cable looms are distributed all around the ITER device and attached to the inner wall of the Vacuum Vessel (VV) and under the divertor cassettes. The results include distributions of neutronics characteristics calculated with the MCNP5 3D Monte Carlo code assuming ITER operation with 0.54 Full Power Years on 500 MW fusion power of DT neutron source defined in the standard ITER B-lite neutronics model. Neutron and photon fluxes, neutron energy group fluences, nuclear heating, neutron damage, and neutron-induced transmutation of the cable looms’ materials have been analyzed in this paper. This work provides data for more comprehensive integral analysis of the cabling system with consideration of thermo-mechanical, electromagnetic, and structural aspects. The temperature gradients along the cable looms and nuclear transmutation of copper could result in formation of thermocouples due to the Radiation-Induced Thermoelectric Sensitivity (RITES) effect, which should be reduced in order to enhance the performance of the magnetic sensors. Therefore, reliable thermal contact should be established in the places of attachments of cable looms to the VV inner wall and the divertor cassette body. [ABSTRACT FROM AUTHOR]

• The load under experimental test conditions (including liquid pressure, nitrogen baking), normal operation conditions (including the impact of thermal stress under thermal load, gravity, working pressure, etc.) and extreme event conditions (VDE event or MD event, including the impact of thermal stress under thermal load, gravity, working pressure, electromagnetic load, etc.) of the OVT are simulated and studied. • The structural strength of the OVT meets the design requirements under different working conditions, and the OVT has good mechanical properties. • The analysis results under various working conditions of the OVT are summarized. When the OVT is subjected to thermal loads, the impact of thermal stress caused by the structural temperature gradient and the difference in the thermal expansion coefficients of various materials puts forward higher requirements for the mechanical properties of the OVT, and the impact degree is greater than the electromagnetic load, experimental test load, and occupies a dominant position. • The work completed will promote the processing and manufacturing of the OVT prototype. China Fusion Engineering Test Reactor (CFETR) is an important device for studying fusion reactor engineering in China. The Outer Vertical Target (OVT) are high heat load components of the divertors and work in a strong electromagnetic field, so they need sufficient heat removal capability and structural strength. In addition, due to activation of neutrons on materials, the OVT also needs to meet the requirements of Remote Handling (RH) maintenance. CFETR puts forward the concept of hybrid divertor-blanket, which cancels the cassette and adopts a separate structure. The lower part of the divertor is occupied by the breeding blanket, and the divertor target is fixed on the breeding blanket, to improve the RH maintenance efficiency and Tritium Breeding Ratio (TBR). In previous work, the structural design and thermal-hydraulic analysis of the OVT were completed. In this paper, load study and structural analysis was carried out to verify the structural strength of the OVT under experimental test conditions (including liquid pressure, nitrogen baking), normal operation conditions (including the impact of thermal stress under thermal load, gravity, working pressure, etc.) and extreme event conditions (Vertical Displacement Event (VDE) or Major Disruption (MD) event, including the impact of thermal stress under thermal load, gravity, working pressure, electromagnetic load, etc.). The results show that the OVT meets the requirements of heat removal and strength, and effectively promote the processing and manufacturing of the prototype of the OVT. [ABSTRACT FROM AUTHOR]

Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained. [Copyright &y& Elsevier]

Abstract: This paper describes the eddy current computation and the resultant forces and torques on selected shield modules assigned to the US team that occur due to plasma disruption. The plasma disruption considered is referred to as major disruption (MD) and is one of the disruption cases defined by the International Organization (IO). This paper identifies the applicability of geometrical simplifications for future design analyses. In particular it is shown that cutting a module in half does not preserve the physics of the eddy current generation and resultant calculations while modeling a full module including the nearest modules does preserve the fundamental physics. The force results are shown for shield modules 7 and 13 exposing the validity of geometrical simplifications. The computed torque for these two modules is also presented. [Copyright &y& Elsevier]

The interaction cavity of the European 170 GHz, 1 MW Continuous Wave (CW) gyrotron for ITER, which could also be water-cooled using mini-channels as recently proposed, experiences during operation a very large heat load (>15 MW/m 2 ) localized on a very short (<1 cm) axial length. Such heat loads are typical for high power gyrotrons. As the thermal deformation of the cavity influences the electromagnetic field structure and consequently the gyrotron operation, the analysis of the cavity performance requires the simultaneous solution of the coupled thermal-hydraulic, thermo-mechanic and electro-magnetic fields. In this paper, the thermal behaviour of the cavity under nominal heat load is computed first by CFD. Then a 3D thermo-mechanical model of the cavity is developed, based on the temperature maps computed by CFD, to evaluate the resulting deformation of the inner cavity surface. Finally the deformation is used to compute the updated heat load coming from the electromagnetic field generated by the electron beam in the deformed cavity, which becomes the input for a new iteration of the thermal-hydraulic, thermal-mechanical and electromagnetic analyses. It is shown that this iterative procedure converges to a self-consistent heat-load/temperature-field/deformation-field picture in nominal operating conditions, without exceeding a temperature of ∼230 °C on the inner surface of the cavity. [ABSTRACT FROM AUTHOR]

International Thermonuclear Experimental Reactor (ITER) is a full superconducting coil tokamak. The strap composed of fiber glass composite materials is an important component of Poloidal Field (PF) coil tail. The main function of the strap is to maintain the electrical isolation and mechanical connection between the last turn and the last but one turn. The thermal contraction of the strap shall be measured to assess the mismatch between the strap and adjacent conductor while the coil is cooled down, and further, in order to balance the thermal stresses, pretension on the strap is needed while the elastic modulus of the full-size strap shall also be measured. In addition, the fatigue performance of the strap shall be examined in consideration of the cyclic electromagnetic load on the tail in the tokamak operation condition. In this paper, the thermal contraction test, the elastic modulus test and the fatigue test on the full-size strap for ITER PF coils were carried out. The test results show that the thermal contraction from 300 K to 77 K is 0.19% and the elastic modules is about 60 GPa at 77 K. The fatigue results indicate that the strap can meet the requirement of ITER and can be utilized to the practical applications. [ABSTRACT FROM AUTHOR]

A Helium Cooled Solid Blanket (HCSB) for CFETR (Chinese Fusion Engineering Test Reactor) was designed by USTC. The structural and thermal-hydraulic analysis has been carried out, while electromagnetic analysis was not carefully researched. In this paper, a FEM (finite element method) model of the HCSB was developed and electromagnetic forces as well as moments was computed by a FEM software called MAXWELL integrated in ANSYS Workbench. In the geometrical model, flow channels and small connecting parts were neglected because of the extreme complication and the reasonable conservative assumption by neglecting these circumstantial details. As for electromagnetic (EM) analysis, Lorentz forces due to eddy currents caused by main disruption and Maxwell forces due to the magnetization of RAFM steel (i.e. EUROFER97) were computed. Since the unavailability of the details of the plasma in CFETR, when disruptions happen, the condition where a linear current quench of main disruption occurs was assumed. The maximum magnitude of the electromagnetic forces was 356.45 kN and the maximum value of the coupled electromagnetic moments was 1899.40 N m around the radial direction. It is feasible to couple electromagnetic analysis, structural analysis and thermal-hydraulic analysis in the future since MAXWELL has good channels to exchange data between different analytic parts. [ABSTRACT FROM AUTHOR]

The performance of Nb 3 Sn cable-in-conduit-conductor (CICC) shows a significant degradation with increasing electromagnetic load in the International Thermonuclear Experimental Reactor (ITER). As the strand is under the compression caused by adjacent strands, the strand damage may occur for high contact force. In this paper, we present a 3D helix model based on the TEMLOP and FEMCAM to simulate the contact force among the strands under transverse load. The maximum linear strain and indentation depth induced by contact stress are calculated with two-dimensional contact model of cylinder. The numerical results are compared with the experimental values. Finally, the indentation depth is discussed for different layers and cross angles based on two models. With the increasing of layer number, the degradation of performance will occur. The short bending wavelength leads to small indentation depth. [ABSTRACT FROM AUTHOR]

The Test Blanket Module (TBM) will be used in the test port of ITER to demonstrate tritium self-sufficiency and the extraction of high grade heat for electricity production. J-TEXT TBM mock-up using reduced activation ferritic/martensitic (RAFM) steel as structural material was designed and fabricated to perform and validate relevant electromagnetic and thermal technologies of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) on the J-TEXT. Its size is one third of the CN HCCB-TBM. By using the finite element analysis technology, this paper analyzed the impacts on the equilibrium magnetic field over the plasma region after introducing the structure material RAFM steel. The distribution of toroidal field (TF) ripple and the magnitude of the error field with the mock-up at different positions were given. Simulation shows the distribution of the null field region formed by poloidal field (PF). The influence to tokamak discharge has been evaluated by drawing the magnetic field lines. Based on the results above, we have optimized and finished the installation of the mock-up to J-TEXT which meets the needs of the experiments and to ensure the normal discharge. [ABSTRACT FROM AUTHOR]

Korean helium cooled ceramic reflector (HCCR) test blanket module set (TBM-set) will be installed at equatorial port #18 of Vacuum Vessel in ITER in order to test the breeding blanket performance for forthcoming fusion power plant. Since ITER tokamak has a set of electromagnetic coils (Central Solenoid, Poloidal Field and Toroidal Field coil set) around Vacuum Vessel, the HCCR TBM-set, the TBM and associated shield, is greatly influenced by magnetic field generated by these coils. In the case of fast transient electromagnetic events such as major disruption, vertical displacement event or magnet fast discharge, magnetic field and induced eddy current results in huge electromagnetic load, known as Lorentz load, on the HCCR TBM-set. In addition, the TBM-set experiences electromagnetic load due to magnetization of the structural material not only during the fast transient events but also during normal operation since the HCCR TBM adopts Reduced Activation Ferritic Martensitic (RAFM) steel as a structural material. This is known as Maxwell load which includes Lorentz load as well as load due to magnetization of structure material. This paper presents electromagnetic analysis results for the HCCR TBM-set. For analysis, a 20° sector finite model was constructed considering ITER configuration such as Vacuum Vessel, ITER shield blankets, Central Solenoid, Poloidal Field, Toroidal Field coil set as well as the HCCR TBM-set. Three major disruptions (operational event, likely event and highly unlikely event) were selected for analysis based on the load specifications. ANSYS-EMAG was used as a calculation tool. The results of EM analysis will be used as input data for the structural analysis. [ABSTRACT FROM AUTHOR]

In SST-1, successful plasma startup has been achieved at very low loop voltage with the help of ECRH pre-ionization. ECRH is operated in both O mode and X mode for the purpose of pre-ionization at the pre fill pressure of 1 × 10 −5 mbar. A delay in breakdown has been observed in case of second harmonic ECRH pre-ionization; where in case of fundamental mode of ECRH pre-ionization, the instant breakdown has been observed. This work has attempted at explaining the non-linear interaction of the seed electrons with the electromagnetic field of the incident ECRH wave that has led to break down of the plasma. The delay in the break down attributes to the time differential between the applications of the ECRH pulse to that of the appearance of the H-alpha signal in SST-1. The observed experimental results have been discussed in this paper from the first principles and numerically solving the electron-ECRH field interactions resulting in energy gains of the electrons leading to plasma break down in SST-1 specific discharge conditions. [ABSTRACT FROM AUTHOR]

Faraday shield is a thin copper structure with a large number of slits which is usually used in inductive coupled plasma (ICP) sources. RF power is coupled into the plasma through these slits, therefore Faraday shield plays an important role in ICP discharge. However, due to the complex structure of the Faraday shield, the resulted electromagnetic field is quite hard to evaluate. In this paper, a 2D model is proposed on the assumption that the Faraday shield is sufficiently long and the RF coil is uniformly distributed, and the copper is considered as ideal conductor. Under these conditions, the magnetic field inside the source is uniform with only the axial component, while the electric field can be decomposed into a vortex field generated by changing magnetic field together with a gradient field generated by electric charge accumulated on the Faraday shield surface, which can be easily found by solving Laplace's equation. The motion of the electrons in the electromagnetic field is investigated and the results show that the electrons will oscillate along the radial direction when taking no account of collision. This interesting result brings insight into how the RF power couples into the plasma. [ABSTRACT FROM AUTHOR]

The generic upper port plug (GUPP) structure in ITER is a 6 m long metal box which deploys diagnostic components into the vacuum vessel. This structure is commonly used for all the diagnostic upper ports. The final design of the GUPP structure, which has successfully passed the final design review in 2013, is described here. The diagnostic port plug is cantilevered to the vacuum vessel with a heavy payload at the front, so called the diagnostic first wall (DFW) and the diagnostic shield module (DSM). Most of electromagnetic (EM) load (∼80%) occurs in DFW/DSM. Therefore, the mounting design to transfer the EM load from DFW/DSM to the GUPP structure is challenging, which should also comply with thermal expansion and tolerance for assembly and manufacturing. Another key design parameter to be considered is the gap between the port plug and the vacuum vessel port. The gap should be large enough to accommodate the remote handling of the heavy port plug (max. 25 t), the structural deflection due to external loads and machine assembly tolerance. At the same time, the gap should be minimized to stop the neutron streaming according to the ALARA (as low as reasonably achievable) principle. With these design constraints, the GUPP structure should also provide space for diagnostic integration as much as possible. This requirement has led to the single wall structure having the gun-drilled water channels inside the structure. Furthermore, intensive efforts have been made on the manufacturing study including material selection, manufacturing codes and French regulation related to nuclear equipment and safety. All these main design and manufacturing aspects are discussed in this paper, including requirements, interfaces, loads and structural assessment and maintenance. [ABSTRACT FROM AUTHOR]

The Nb 3 Sn strands which have high critical field are used in cable-in-conduit conductors (CICCs). The superconducting strands are twisted multistage and experience complex thermal and electromagnetic loadings. Due to their brittleness, the cracking of the Nb 3 Sn filaments will occur under mechanical loading. In this paper, based on the linear elastic fracture theory, we study the effects of tension loading on the fracture behavior of central crack firstly. The strain energy release rates for different twist pitches and cabling stages are presented. As the triplet is subjected to the uniaxial strain, the cracking probability will increase with the twist pitch. The crack number increases with the applied strain, and wider filament or bronze can lead to smaller crack number under the same applied strain. In addition, multistage cabling has better mechanical stability. Next, the two collinear crack problem is considered. The variations of microcrack number show that the wider bronze can provide more resistance for the propagating of the large cracks. We can conclude that the bronze plays an important role in improving the stability and strength. [ABSTRACT FROM AUTHOR]

In order to fully validate “ITER-like” actively water cooled tungsten plasma facing units, the implementation of an axisymmetric divertor structure in the Tokamak Tore-Supra has been studied. With this major upgrade, the so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the issues of long plasma discharges using a tungsten divertor based on monoblock targets. The divertor structure and coils assembly are made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) in which a total divertor current of up to 16 × 13 kA is circulating in steady state. The conductor is electrically insulated and wedged inside the casing in order to be mechanically protected. The divertor which is designed to perform steady state plasma operation (up to 1000 s), must sustain harsh environmental conditions in terms of ultra light vacuum conditions, electromagnetical loads and electrical insulation (5 kV ground voltage) under high temperature (180 °C). Therefore, a feasibility study of such a complex structure has been performed. It implied activities on a scale one dummy coil, such as installation, assembly issues and representative tests (electric, thermal and hydraulic). The manufacturing of the divertor structure, which is a large assembly of 4-m diameter representing a total weight of around 20 tonnes, started in the second half of 2013 and is expected to be delivered by the end of 2014. The paper will illustrate the technical developments and tests performed during 2013 and beginning of 2014 in order to fully validate the design concept before the industrial phase. The manufacturing methods proposed by the contractor in order to fulfil the technical requirements will be also addressed. [ABSTRACT FROM AUTHOR]

The purpose of this paper is present an overview of the methodology followed to calculate the dynamic amplification factors applied to the electromagnetic loads acting in the ITER generic port plugs. The methodology used for combining an EM transient event with another kind of load is based in the treated of this dynamic EM event as a static load. As first stage a transient dynamic analysis was performed, at the most demanded electromagnetic event [2] , to determine the dynamic response of the port plugs. In the same way, have been solved all the time steps of the dynamic event as static loads, it means that the inertial effect has been neglected. The response of each time-step at the dynamic solution has been compared with the same time-step solved as a static load. For this purpose some control points were positioned along the structure at the most representative locations. The key of these calculations is the understanding of the deformed modes affecting the port plug in order to obtain a reasonable dynamic amplification factors that permit the characterization of these loads in a realistic way and does not derive in a too conservative approach. Additionally, the fundamental frequencies and vibration modes of the generic port plug, requested for the characterization of the damping effects at the structure, were calculated in a complementary modal analysis performed for this aim. [ABSTRACT FROM AUTHOR]

The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor. [ABSTRACT FROM AUTHOR]

The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components, which in turn is related to the amount of mobilised dust present in the Vacuum Vessel. Periodically or on request, the IVVS scanning probes will be deployed into the Vacuum Vessel in order to acquire both visual and metrological data on plasma facing components (blanket, divertor, heating/diagnostic plugs, and test blanket modules). Recent design changes made to the six IVVS port extensions implied the need for a substantial redesign of the IVVS integrated concept – including the scanning probe and its deployment system – in order to bring it to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the concept design for IVVS as well as of the various engineering analyses and R&D activities carried out in support to this design: neutronic, seismic and electromagnetic analyses, probe actuation validation under environmental conditions. [ABSTRACT FROM AUTHOR]

The magnet system of the stellerator Wendelstein 7-X (W7-X) consists of 5 modules of 14 superconducting coils with complex 3D shape each. After manufacturing the coils and assembly of the modules on temporary stands, the position of each module on the machine base was successfully optimized to minimize the electromagnetic (EM) field asymmetry. This asymmetry originates from inevitable geometric deviations of the coils from the target shape due to manufacturing and assembly tolerances. However, new deviations were introduced after module optimization due to bolting the modules of the magnet system together to a torus, removing temporary supports and further loading of the machine base with weight of additional components. In this paper, the geometrical deviations along the centre line of the coil currents are assessed through detailed step-by-step non-linear finite element (FE) simulation of the assembly procedure of the complete torus. The model is evaluated against measured displacements and reaction forces monitored during consequent assembly steps. The results are being used to quantify the obtained field asymmetry and countermeasures to minimize it. [ABSTRACT FROM AUTHOR]

Abstract: The core charge exchange recombination spectroscopy (cCXRS) upper port plug diagnostic system serves to control and optimize plasma performance. The system contains a labyrinth of optical mirrors delivering the visible light from the plasma to the end spectrometers. Harsh ITER near plasma environment, resulting in high thermal fluxes and severe electromagnetic (EM) loads, dictates a necessity of a robust design for the plug components. The structural response of the plug components to combined loading coming from different sources has to be calculated and analyzed. The global ANSYS EM model featuring the reference cCXRS layout has been upgraded to be in line with the recent ITER generic upper port plug. Impact of different parameters on the plug EM loading was studied. The dynamic structural behavior of the plug (new baseline layout) under the EM forces has been analyzed. The plug lateral deflection is about 9mm and does not exceed the ITER limit of 12mm. The cooling and supporting system of the cCXRS first mirrors, located near the plasma, have to cope with significant EM loads and high heat fluxes. The engineering ‘express’ approach, that was recently effectively used to calculate the eddy currents and EM forces for the cCXRS fast shutter, has been further validated for the mirrors. The paper outlines approaches to a multifield analysis of the first mirror. [Copyright &y& Elsevier]

Abstract: The final design of ITER vacuum vessel thermal shield (VVTS), which is planned to be procured completely by Korea, has been implemented after the procurement arrangement was signed. In this paper, the design and the supporting analysis are described for the key components of the VVTS such as joint, panel, support, and stopper. The VVTS design is revised and finalized based on the manufacturing feasibility, interface requirement and assemble feasibility. The inboard and the outboard supports of VVTS are designed in detail considering structural rigidity and assemble feasibility. The shape of in-pit joint, which is installed every 40° sector in toroidal direction for compensation of possible misalignment during sector assembly, is determined. Three types of joints are developed in accordance with their locations and assemble feasibilities are checked through the R&D. Stopper design is developed in order to prevent direct contact against adjacent components such as vacuum vessel and magnets. Structural rigidity of the whole VVTS is also validated by finite element analysis under various kinds of operating conditions, such as deadweight, electro-magnetic load, seismic load and load combinations. [Copyright &y& Elsevier]

Abstract: The enhancements carried out in the tokamak ASDEX Upgrade (AUG) are oriented toward the preparation of the future physics-related activities of ITER and DEMO. To address the main ITER issues, plasma configurations with a wider operational limit (e.g. higher triangularity) are planned for the future experimental campaigns in AUG. To evaluate the mechanical impact on the toroidal field (TF) magnet system a combined electromagnetic and structural finite element model was set up. At first extensive benchmarks of the models are carried out against the AUG reference design configurations with respect to stress [1–3], lateral displacement measurements and poloidal flux pattern. The numerical model was then applied to a set of actual high triangularity (HT) configurations generated by a more favorable poloidal field (PF) current distribution made possible by an extension of the power supply system. The resulting change of the poloidal flux pattern and the lateral force distribution has consequences for the coil shear stress and vault stability. Both aspects are monitored by a safety system measuring the PF flux placed on top and bottom of the outer surface of two TF coils (TFCs) between vault and the TFC supporting structure, so called Turn Over Structure (TOS). The range of the new HT configurations has induced a modification of the flux pattern, so that an adaptation of safety system is required to protect the TFCs system. Following the same criteria of the old safety system [4,5], a new set up of virtual coils will be integrated in the control system of AUG. The motivation of this new set up will be discussed in this paper. [Copyright &y& Elsevier]

Abstract: The aim of the PFX Operating Early Task (POET) was to obtain a highly shaped plasma and x-point formation in the early phases of the discharge. The PFX is an amplifier which feeds the central pancakes of the JET primary circuit. In the past it was possible to energize the PFX circuit only when the PFGC current, which feeds all the coils of the primary circuit, was already flowing in the same direction as the PFX current would have flowed, to avoid repulsive vertical forces which would tend to lift the top pancakes of the central solenoid, balanced only by the net weight of the upper part of the JET machine. In this paper the modeling activity performed to provide a more accurate estimate of the ejection forces acting on the upper coils in order to safely widen the operational space, by using two dimensional finite element electromagnetic models and the simulation of the performances of the actual controller algorithm on tracking the desired references of current are described. Finally will be presented the results of the implemented POET system, routinely used in JET in the 2011/2012 experimental campaigns, in terms of anticipation of x-point formation and enhanced system flexibility. [Copyright &y& Elsevier]

Abstract: The distribution of forces, acting on components of the electromagnetic (EM) system considered herein has been analyzed with the use of simplified models and equivalent dynamic simulators for a variety of plasma event scenarios. The work was aimed at the formulation of the problem of the global force balance and the approach to its resolution, identification of the main factors of the force interaction between components, development of the numerical models for qualitative and quantitative assessment of the problem and relevant parameters and the computations and assessment of the results. The use of equivalent simulators with elimination of redundant details and degrees of freedom (DOFs) principally reduces computational complexity of the problem. The selection (identification) of the equivalent system on the basis of the evaluation of the significance of impacts on the parameter of interest provides the effective and natural filtering of secondary or irrelevant information. The finding of equivalent minimum dynamic systems (EMDS) opens, at a certain rate of computations, a way to develop real-time simulators. The paper describes simplified models, the principles of definition (synthesis) and use of EMDS, the structure of the database (magnitude and time evolution of distributed and integral basic parameters) required for the evaluation of the EM loads and examples of the data. [Copyright &y& Elsevier]

Abstract: Japan Atomic Energy Agency (JAEA), acting as the Japan Domestic Agency (JADA) in the ITER project is responsible for the procurement of 9 Toroidal Field (TF) coil winding packs (WPs), structures for 19 TF coils (including one spare), and assembly of the WPs and the coil structures for 9 TF coils . The TF Coil structures which support large electromagnetic force generated in TF coils under the cryogenic temperature (about 4K), are very large welded structures composed of a coil case and structural attachments made of high strength and high toughness stainless steel. JAEA has been performing welding trials for heavy thickness materials since 2008 and is planning fabrication of full-scale mock-ups of main sub-components (1 set for the inboard side and 1 set for the outboard side) in 2011 in order to investigate the technical issues for manufacturing the TF coil structures. This paper presents the results of welding trials and the status of full scale mock-ups fabrication to confirm the validity of welding technology and manufacturing design before fabricating the actual products. [Copyright &y& Elsevier]

Abstract: Four of the 16 ITER upper port plugs will be devoted to electron cyclotron resonance heating (ECRH) in order to control magnetohydrodynamic (MHD) instabilities . In order to achieve the stabilisation of the neoclassical tearing modes (NTM) and sawtooth oscillation, a deposition of a very localized and peaked current density profile over a broad poloidal steering range is required. In the present optical configuration eight 2MW mm-wave beams enter each of the four upper launchers (UL) through waveguides into the vacuum vessel. Each beam line comprises consecutive corrugated waveguide sections with two mitre bends, orientating the poloidal and toroidal directions and three sections of quasi-optical transmission . The beam waist locations and beam shaping properties in free space propagation are defined by two additional mirrors, the first being a static focusing mirror and the second a plane poloidally steerable mirror. Each mirror reflects a group of 4mm-wave beams. The three types of UL mirrors (mitre bend, focusing and steering) absorb heat generated essentially by three sources: the ohmic loss of the RF beam reflected at the mirror surfaces and the nuclear and thermal radiation coming from the plasma. While the average heat load is within reasonable engineering limits, three elements condition the actual mirror design, the peak ohmic heat load (Gaussian or Bessel type heat deposition profiles), the electromagnetic forces generated in vertical disruption events (VDE), and the ITER cooling water requirements. This paper provides an overview of the different upper port-plug mirror designs and cooling schemes and an outlook on the prototype manufacturing activities and the future test program. The optimized mm-wave layout within the ECH port plugs is also presented. [Copyright &y& Elsevier]

Abstract: The use of thin liquid metal layers as shield against erosion and thermal loads of the inner reactor walls continues to be of interest. The degree and reliability of the protection depend on the stability of the liquid metal layer under the influence of the surrounding magnetic field. In this paper we treat analytically the problem dealing in particular with the stability of the liquid metal layer under the influence of a uniform surface normal magnetic field and its control by means of a transversal electrical field. The Karman–Polhausen integral method is used to reduce the system into two equations that describe the spatio-temporal variations of the flow rate and the film thickness. The proposed model improves the formulation of Korsunsky (1999) for high Reynolds numbers by including the second order terms. It is shown that Korsunsky''s formulation overestimates the cut-off frequency and it does not foresee periodic stationary waves to develop on the free liquid surface. The electrical and magnetic fields’ abilities to stabilize or destabilize the flow, as a function of the critical Reynolds number, are examined. The influence of the electromagnetic fields on the cut-off wave-number and growth rate of the perturbations in terms of the Reynolds number and wave-number is also presented. [Copyright &y& Elsevier]

Abstract: The paper describes the studies and the tests for the development of the insulation structure of the 1MV–50A gas insulated (SF6) line of the ITER NBI in the SinGap configuration characterized by two kinds of spacers: at least a couple of disk-shaped spacers, designed to be gas tight, and a larger number (several tens) of inner conductor post spacers. To this aim a test campaign has been carried out to assess the capability of standard epoxy spacers to withstand a high dc voltage with frequent short circuits, simulating the operational condition for the ITER NBI. Two computational tools, the first for the epoxy spacer shape optimization under electrostatic distribution and the other for the nonlinear time variant evolution of the electric field and surface charge, have been developed specifically for designing epoxy spacer under dc voltage stress. The results on the optimization of the disk spacer and on the electric field–surface charge time evolution of the post spacer are reported and discussed. The effects of the SF6 radiation induced conductivity on the post spacer are also reported. [Copyright &y& Elsevier]

Abstract: The main ITER cryogenic users are the superconducting magnets together with the 80K thermal shields. The key requirements that drive the design of the ITER cryogenic system are the intrinsically pulsed heat deposition in the magnets due to electromagnetic field and nuclear heating cyclic variations, and the usage of large flow capacity helium circulating pumps for forced-flow cooling of the magnets. This paper highlights the main features of the ITER cryogenic system, including the results of studies to mitigate pulsed heat loads on the helium refrigerator and minimize the heat load from helium circulating pumps. [Copyright &y& Elsevier]

Abstract: The current decay behaviors in high-temperature superconducting (HTS) coils using Bi-2223/Ag tapes are investigated. Generally, the resistance of a superconducting closed loop is determined by the joints and electric field versus current density relationship of the HTS tape. In this paper, we demonstrate that the resistance is enhanced due to the decay of the shielding current by the flux creep and flux flow in HTS tapes. Experiments with HTS coils with persistent current operations were carried out and the results are compared with numerical analysis using a finite element method. Furthermore, the improvement of the current decay behavior is also discussed by choosing an overshooting excitation technique. [Copyright &y& Elsevier]

The paper presents the various aspects of the new set of 192 saddle coils (SC), designed for the active control of MHD plasma modes in the RFX experiment. The choices necessary to fulfil the electromagnetic specifications and the geometrical constraints of the system are described. The methods and results of thermal, electrodynamic and mechanical analyses carried out for the proper design of the coils are also reported. The technological solutions proposed for the construction, presently in progress, are illustrated. [Copyright &y& Elsevier]