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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.

The latest achievements in the development of high performance Nb3Sn strands are reported. In the field of high energy physics (HEP) critical current densities up to 3000 A mm−2 (at 12 T, 4.2 K and 10 µV m−1) can be achieved. Current actions are aimed at reducing the filament size to about 50 µm and decreasing production costs by using less expensive base materials. In order to take advantage of the progress made by the HEP R&D program, the fusion community has initiated several strand procurement actions worldwide to produce advanced Nb3Sn strand types fulfilling the updated strand specification of the International Thermonuclear Experimental Reactor project. Recently developed internal-tin strands reach critical current densities of more than 1100 A mm−2 and acceptable non-Cu hysteresis losses of about 700 kJ m−3. An important milestone towards higher production rate and yield was reached: a billet of about 100 kg was successfully drawn to the final strand diameter without problems. In powder-in-tube conductors, the composition and number of filaments have been modified to meet the European target specification. Although the bronze technique itself sets certain limitations on Jc, strands with critical current densities around 800 A mm−2 have been produced on an industrial scale.

Strands relevant for fusion with high critical current densities and moderate hysteresis losses were developed and already produced on industrial scale. Based on these achievements EFDA-CSU Garching has launched a Nb 3 Sn strand development and procurement action inside Europe in order to assess the current status of the Nb 3 Sn strand production capability. All six addressed companies have replied positively to the strand R&D programme which includes the three major Nb 3 Sn production techniques namely the bronze, internal-tin and powder-in-tube (PIT) route. According to the strand requirements for the ITER TF conductor a critical current density of 800 A/mm 2 (at 12 T, 4.2 K and 10 μV/m) and overall strand hysteresis losses below 500 kJ/m 3 have been specified as the minimum guaranteed strand performance. The second major objective of this programme is to motivate the strand manufacturers to develop and design high performance Nb 3 Sn strands optimised for the ITER conductor. For this purpose, a target critical current density of 1100 A/mm 2 has been added to the specification. This paper describes the strategy behind the strand development programme, the actual status of the strand production as well as first preliminary results obtained from the strand suppliers.

The paper debates the Toroidal Field Model Coil (TFMC) main design and fabrication choices, as well as, coil fabrication issues with respect to technological and project management aspects. The latest test result and the experience gained during the assembly of the TFMC will also be presented. Finally the next major milestones in coil testing and performance analysis are presented.

(1998). ITER Toroidal Field Model Coil (TFMC) Design and Construction. Fusion Technology: Vol. 34, Proceedings of the Thirteenth Topical Meeting on the Technology of Fusion Energy. Nashville, Tenessee. June 7–11, 1998, pp. 797-801.

Objective of the Safety and Environmental Assessment of Fusion Power (SEAFP) [1] is to `assess the safety and environmental features of fusion on the basis of the conceptual studies of tokamak power reactors which incorporate safety and environmental features likely to ensure public acceptance and take into account the operating requirements of utilities'. The fusion reactors for commercial use are expected to enter in operation in several decades, therefore the assessment shall be as relevant as possible with respect to further development in physics and technology. The reactor has been conceived for monitored and filtered release of radioactivity to the environment during normal operation equivalent to a small fraction of the natural exposure; to avoid radioactivity release to the environment for all in-plant accidents and conceivable ex-plant accidents; to avoid disruption of community life outside the plant perimeter even in the case of rupture of confinement barriers, in particular by non-anticipated plant initiators. Consideration of possible fault conditions is an integrated part of the design of the main components and their associated support systems. the overall philosophy proposed is that all faults should be contained within an individual system regardless of the initiation fault or external action.

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 aims of the toroidal field (TF) model coil are to test the manufacturing feasibility of the ITER TF magnet concepts, to assess the reliability of the fully integrated system by dedicated testing and to qualify the quality assurance and the testing methods. The 3.8 m long racetrack shaped TFMC will be tested in the TOSKA facility at Karlsruhe in a configuration using the Euratom LCT Coil to provide in-plane and out-of-plane loads, relevant to the ITER coils.

Several prototype Cable‐In‐Conduit‐Conductors (CICC) for the superconducting EDIPO (Efda DIPOle) revealed a degradation of their critical current (Ic) increasing with each loading cycle. The strong Lorentz‐forces during operation in combination with the limited support of the single strands against these forces are thought to be the cause of the permanent degradation of the brittle Nb3Sn superconductor from which the multi‐stranded CICC are made. In summer 2006 EFDA started to explore the possibility to remedy the Ic degradation by solder‐filling the conductor in order to mechanically stabilize the twisted‐strand cable inside the conduit. This solution was not considered as the main one, but as an emergency solution to be applied to the completed magnet, should every other option fail. The solder‐filling approach was previously applied with success in some cases. Some issues, however, needed to be clarified before this solution could be proposed for the EDIPO project. The most important among them are the choice of solder material, details of the solder filling process, and the thermo‐mechanical implications of a solder‐filled, high‐field, high‐current cable. This work, being reported here, made use not only of simulation but also of experiments, such as the mechanical testing of solder filled cables at cryogenic temperatures.

The European Fusion Programme included since the mid-1990s, the development of a full scale, full rating current commutating unit (CCU) to be used as a centre-piece of the ITER magnet protective circuits. The CCU is mainly composed of a by-pass switch (BPS) in parallel to a vacuum circuit breaker (VCB). Both full scale switches were subjected to extensive testing in conditions simulating the actual ITER operation. The operation in absence of the saturable reactor series connected in the ITER reference design was successfully assessed. Interruption tests at low current and full counter-pulse capacitor voltage, were also satisfactorily performed. The tests included an extensive and novel characterisation of the interrupting capability of the vacuum circuit breaker in the presence of an external magnetic field as in the actual location of installation in ITER.

In the framework of the activities for the CS Model Coil of ITER, the European Union (EU) is in charge of providing 6.5t of bronze Nb3Sn strand, cabling for 798m of MCI and 396m of MC2 cables and final conductor manufacture (jacketing and compaction) for all the conductor lengths of the CS Model Coil, for a total of 5787m. The related industrial activities are in progress. This paper describes in some details the manufacturing experience gained in EU and the present status.

The aim of this paper is to present an up to date review of the status of the European DIPOle (EDIPO) project-whose objective is to build a new facility to perform both DC and AC tests of large force flow superconductor samples in high magnetic background field (up to similar to 12.5 T)-as well as an outlook on the final commissioning of the magnet in CRPP-PSI foreseen for 2012. Concerning the actual status, the results presented focus on the recent steps carried out in the manufacturing of the magnet assembly, in particular the cold mass assembly, its impregnation and the final acceptance tests. Moreover, results are reported regarding the identification and repair of an inter-turn short carried out on EDIPO Pole 1 in the second half of 2010. As far as the commissioning of EDIPO is concerned, an up-to-date report is provided on the preparation of the facility at CRPP (e.g. mechanical structures, high temperature current leads, quench detection, control system, etc.) as well as its acceptance and final commissioning (e.g. magnet reception tests, commissioning program, etc.)

Hadron physics has been an essential part of the physics program with the CLAS detector in experimental hall B at Thomas Jefferson National Accelerator Facility (Jefferson Lab). With the 12 GeV upgrade of the CEBAF machine, hadron physics in Hall B will be extended to a new domain of higher mass resonances and the range of higher transferred momentum using up to 11 GeV electron beams and the upgraded CLAS12 detector. In this paper, status of the hall B superconducting magnets for the CLAS12 detector is presented.

DIOSCUR is a superconducting Tokamak (R=3.7 m, a = 1.05m, kx=2), which at a magnetic field of 7 T can sustain a current of 10 MA. With 50 MW of total power, plasma density and temperature are expected to be close to those of a thermonuclear plasma, so that it would be straighforward to extrapolate the results of DIOSCUR to a reactor. The main goal of the device is to attain steady state operation by solving the problems connected with the handling of heat and particle fluxes on the first walls and an efficient current drive (CD) system. A large degree of flexibility has been emphasized in the DIOSCUR design: this will permit to investigate plasma regimes not yet fully explored in present experiments. Thus DIOSCUR could be both of support and complementary to the ITER programme.

An alternative to the basic NET first wall (FW). concept is presented. The concept aims at: minimizing the maximum surface temperature of the protection material, the impurity influx from the wall, the intervention frequency for maintenance as well as increasing the operating margins against the heat loads. This will allow to reduce: the required expansion volumes in case of accidental water ingress in the plasma chamber and the tritium inventory present in the protection material. The solution presented is made of removable local limiters (LL) and the proper first wall in between. The main role of the local limiters is to intercept the particle flux during burn (α particle), during start up when they act as plasma limiters and during disruption to stop the run away electrons. The technologies envisaged to make the local limiter are already being developed for the divertor or they could be a limited extrapolation of existing one. The FW will be coated with methods already existing (e.g. plasma spray or physical vapor deposition), but they should be qualified for the operating condition of the first wall and in-situ repair. The coating can be directly applied on the blanket box or on tubes which afterwards are assembled on the blanket box.

The Tokamak reactor will be one of the most complex machines man has ever built. The research and development needed to demonstrate its feasibility encompasses the areas of physics, technology and engineering and the solution will be a good compromise between the requirements of these three different fields. It is not advisable to give priority to the needs of any one of them because good performance in all three areas is essential for the realization of a successful fusion reactor.

Current designs for tokamak based fusion reactors usually consider deuterium-tritium reactions with a 50–50% mixture of D-T, to give a maximum power density, although design studies for pure D-D, He3-catalysed and T-catalysed tokamak fusion reactors have been performed. The letter considers a reactor burning a D-T mixture with an enhanced percentage of deuterium to increase the possibility of D-D reactions (a so-called T-catalysed D reactor). Most of the energy needed to maintain the burning plasma comes from D-T reactions, but the D-D reaction results in the generation of extra neutrons (or equally useful tritons or 3He nuclei), which can contribute to the overall neutron economy of the reactor. – A simple plasma energy balance for an ignited tokamak is used to study the usefulness of D-D reactions, with confinement times, safety factors, and beta limits as predicted by existing experimental devices. The authors consider both the improvements necessary in present plasma performance limits to produce useful D-D reaction rates as well as the effect of a lower plasma energy density on reactor size. – The object of the study has not been to consider the economics of such types of reactor, but rather to assess the potential of the D-D reaction to introduce some flexibility into a fusion reactor design. It is evident from design studies on next generation tokamaks that the overall neutron economy will be a major challenge in a fusion reactor and that the reduction of the necessary blanket breeding ratio could improve the feasibility of the overall design.

The magnetic system of NET (Next European Torus) provides a framework inside which ignited plasmas can be confined while nuclear component testing is performed. The toroidal and poloidal field coils are superconducting, since the demonstration of the applicability of this technology to a fusion reactor is an important part of the NET objectives. The toroidal field coils' winding pack is made of double pancakes realized by winding the conductor with dry glass in the final shape. The poloidal field coils are divided into two subsets, the central solenoid stack and the outer equilibrium coils. Superficially both subsets are similar in that both will consist of a 40-kA force flow. He-cooled superconducting conductor in a glass-fiber-reinforced epoxy matrix, and perform both equilibrium and transformer functions. However, because of different operating conditions (field and stress levels), it is expected that the coils will use different conductor designs. >

The NET reference operating scenario (22 s start up, 450 s burn) is derived. The maximum powers required to and from the coil system are 500 MW and 3500 MW, respectively. Further optimisation of the scenario reduces the peak input power to 200 MW. Power supply circuits are proposed which allow the reference scenario to be achieved with a minimum of equipment (1.7 GVA of convertors, 2.6 GW of resistors).

The coil system of the NET machine provides a framework inside which ignited plasmas can be confined while nuclear component testing is performed. The main coils will be superconducting, since the demonstration of the applicability of this technology to ignited tokamaks is an important part of the NET objectives. This report will look at the factors driving the design of the magnetic system and the solutions that have been adopted.

The mechanical structure of JET device, shaped like a large spherical shell is made up of 64 castings with weights varying between 5 to 10 tonnes each. These castings have to withstand the very large forces arising from the toroidal and poloidal magnetic field of the JET machine. An extensive study was carried out to select a suitable material to produce such big castings and to set up a reliable industrial production method to suit the design requirements. The assessment led to the selection of an austenitic nodular cast iron. A review of the static and dynamic mechanical tests, fracture mechanic tests and functional tests carried out throughout all stages during the development and production is presented in this paper.

In order to make the most efficient use of the space available inside the toroidal field coils, the NET bulk shield is integrated into the vacuum vessel. In this way advantage can be taken of the structural properties of the material used to provide the basic shielding for the superconducting coils.

The aim of this paper is to present an up to date review of the status of the European DIPOle (EDIPO) project whose objective is to build a new facility to perform both DC and AC tests of large superconductor samples in high magnetic background field (up to similar to 12.5 T). EDIPO was designed by EFDA, is being built by BNG (Babcock-Noell) and, once completed, it shall be installed near the existing SULTAN facility at the Paul-Scherer-Institute, (CRPP, Villigen, Switzerland). The results presented focus on the manufacture of the dummy coil (which is a 1: 1 model of one of the EDIPO poles) that was completed in spring 2009. Details are given about the dummy coil winding results, coil reaction heat treatment, impregnation and acceptance tests. Moreover, a progress status is provided about the manufacturing of the other parts of the EDIPO magnet assembly, in particular with respect to the manufacturing of the test-well, AC coils, yoke sheets cutting and assembly, the outer cylinder production and welding (316 LN steel, 35 mm thick) and the coils ends preparation. Some results regarding the R&D studies are also reported about the winding trials carried out to improve the winding accuracy and the numerical simulations made to support coil winding (e.g. conductor spring back).

Raising the JET maximum toroidal field from 3.45 to 4.0 T results in increased forces and stresses in the key machine components during normal operations and disruptions. These forces and stresses have been predicted for 4.0 T operation and compared with allowable values derived from tests on used and spare JET coils. In this assessment significant lessons related to the design, manufacture and operation of key Tokamak components have also been learnt. (C) 2001 The European Commission. Published by Elsevier Science Ltd. All rights reserved.

The ITER toroidal field model coil (TFMC) was designed, constructed and tested by the European Home Team in the framework of the ITER research and development program of the Engineering Design Activities (EDA). The project was performed under the leadership of European Fusion Development Activity/Close Support Unit (EFDA/CSU), Ciarching. in collaboration with the European superconductor laboratories and the European industry. The TFMC wits developed and constructed in collaboration with the European industry consortium (AGAN) and Europa Metalli LMI supplied the conductor, The TFMC was tested in the test phase I as single coil and in phase 11 in the background field of the EURATOM LCT coil in the TOSKA facility of the Forschungszentrum Karlsruhe. In phase 1, the TFMC achieved an ITER TF coil relevant current of about 80kA and further representative test results before the end of the EDA. In the more complex test phase [I. the coil was exposed to ITER TF coil relevant mechanical stresses in the winding pack and case. The tests confirmed that engineering design principles and manufacturing procedures are sound and suitable for the ITER TF full size coils. The electromagnetic. thermo hydraulic and mechanical operation parameters agree well with predictions. The achieved Lorentz force on the conductor was about 800 kN/m. That has been equivalent to the Lorentz forces in ITER TF coils. (c) 2005 Elsevier B.V. All fights reserved.

In the ITER superconducting magnet system, operation currents up to 68 kA are required that have to be transferred from room temperature (RT) to 4.5 K by specially designed current leads (CL). The ohmic losses and the thermal conduction of conventional CL cause high heat loads to the refrigeration system. This load can be reduced drastically by the use of high temperature superconductor (HTS) current leads because the superconducting HTS part of this CL causes no ohmic losses and very poor thermal conduction. The Forschungszentrum Karlsruhe and the CRPP Villigen have designed and built a 70 kA current lead using Bi-2223 HTS superconductor in the frame of the European Fusion Technology Programme. This HTS CL was installed and tested in the TOSKA facility of the Forschungszentrum Karlsruhe showing that the current of 68 kA can be carried by the CL even when the highest temperature in the HTS part is 85 K. The CL shows a large safety margin in case of a loss of coolant and even cooling with LN2 has been demonstrated. The option that all ohmic losses in the temperature range between 4.5 K and approx. 80 K can be eliminated using HTS CL is very attractive for ITER because savings in investment and operation costs are possible by removing this load from the cryogenic system. The consequences with respect to investment and operation costs are discussed comparing the ITER design with and without HTS CL which demonstrates that HTS current leads are very attractive and should be used for ITER

Benign familial infantile epilepsy (BFIE) is the most genetically heterogeneous phenotype among early-onset familial infantile epilepsies. It has an autosomal dominant inheritance pattern with incomplete penetrance. Although PRRT2 is the most mutated gene detected in families with BFIE, other mutations in KCNQ2, SCN2A, and GABRA6 genes have also been described. To date, KCNQ3 mutations have been detected in only four patients with BFIE. Here, we describe the clinical pattern and course of an additional individual with BFIE associated with a novel missense heterozygous KCNQ3 c.1850G>C variant inherited by his unaffected father. The incidence of KCNQ3 mutations among BFIE patients is reported to be low in the literature, however, whether this is underestimated is unclear as not all current epilepsy gene panels include KCNQ3.

Bus Bar III (BBIII) is a curvilinear NbTi full-size ITER-type multistrand superconducting cable 7 m long, tested in 2004 at Forschungszentrum Karlsruhe, Germany. The sample was cooled with a forced helium flow and was equipped with many electromagnetic and thermal-hydraulic sensors, to acquire as wide as possible an experimental database, aimed at getting a benchmark for the validation of the THELMA code, a computer tool for the analysis of CICC prototypes and magnets, recently developed and presently under validation. This code can compute the current distribution among the macrostrands used to represent the cable, taking into account their superconducting behavior, as well as the temperature and the magnetic field distribution. The paper deals with the application of THELMA to some tests on BBIII useful for the electromagnetic model validation and presents in detail the analysis setting-up, the results, its critical aspects and limits. A comparison is shown between the measured and the computed waveforms of the electromagnetic data.

The Bus Bar III (BBIII), fabricated within the Toroidal Field Model Coil Task of the International Thermonuclear Experimental Reactor (ITER), was tested at the Forschungszentrum Karlsruhe, Germany, in the spring of 2004. The BBIII consists of an approximately 7 m long NbTi dual-channel conductor with a thick square stainless steel jacket, cooled by forced flow supercritical He. It was energized with currents up to 80 kA and operates in its self magnetic field (up to /spl sim/0.8 T). The BBIII was instrumented with Hall-probe heads and arrays, voltage rings and longitudinal voltage taps for electro-magnetic measurements, in order to get experimental data to be used for the validation of a recently developed hybrid thermal-hydraulic electro-magnetic code (THELMA), as well as for the assessment of the possibility of performing a reliable reconstruction of the current distribution in the conductor cross section under controlled conditions. In the tests, current ramps at different rates were applied to characterize the conductor time constants, while two different resistive heaters (one upstream of the BBIII inlet, another one directly on the BBIII jacket) were separately operated in order to approach current sharing in the conductor and to observe the related current re-distribution. In this paper, a summary of the collected experimental results is presented, with particular emphasis on those aspects more relevant for the forthcoming THELMA analysis.

A computational tool is being developed for the analysis of superconductive magnets, combining detailed descriptions of termination joints and cables to thermo-hydraulic (TH) models. In parallel, an experiment (Stability Experiment Upgrade*/SexUp) has been designed with the target to study the current distribution in cable in conduit conductors (CICC). Finally, the establishment of a reliable method for the measurement of the current distribution profile on the cable cross section is being implemented on the ITER toroidal field model coil (TFMC).

As the test of the PFC1 is foreseen in 2006 at JAERI Naka, Japan, it is essential to consider in detail the lessons learned from the short NbTi sample tests, as well as the issues left open after them, in order to develop a suitable test program of the PFC1 aimed at bridging the extrapolation gap between measured strand and future PF coil performance. Here we consider in particular the following issues: 1) the actual possibility to quench the PFCI conductor in the T-CS tests before quenching the intermediate joint, 2) the question of the so-called sudden or premature quench, based on SULTAN sample results, applying a recently developed multi-solid and multi-channel extension of the Mithrandir code to a short sample analysis; 3) the feasibility of the AC losses calorimetry in the PFCI.

In this paper, we discuss the predictive analysis performed in support of the test program of the International Thermonuclear Experimental Reactor (ITER) poloidal field conductor insert (PFCI). A subset of the test program items was considered, with particular emphasis on DC performance and AC losses. The results and implications of the comparison of selected predictions from different laboratories will be presented.