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In the International Thermonuclear Experimental Reactor, the nuclear radiation escaping from the vacuum vessel reaches the superconducting toroidal field (TF) magnets, affecting the temperature margin \Delta Tmar, that is, the difference between the current sharing temperature and the operating temperature. The TF magnets are designed to operate at a minimum margin \Delta Tmar^\min = 0.7\ \rm K. Recent design activity on in-vessel components, for example, blanket, in-vessel coils for plasma stability, suggests a potential enhancement of the nuclear heat load, leading to a reduction of \Delta Tmar, which is accurately assessed in the paper using the validated 4C code. For the case when the margin goes below the minimum, different possible mitigation strategies are investigated: the first considers the possible reduction of the He bath temperature from the nominal 4.3 K down to 3.8 K, and is proven to be successful. The others consider the possible increase of the dwell time between plasma pulses, and is shown to be inadequate, or the decrease of plasma pulse duration, which turns out to be effective below 300 s. [ABSTRACT FROM PUBLISHER]

The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Science (ASIPP) by incorporating the results of mechanical and thermal analyses as well as the system integration and assembly tolerances in the present CAD model. The feeder design is being finalized progressively, and will be delivered to the Chinese Domestic Agent (CNDA) for further procurement arrangement (PA) activities. Pre-PA manufacturing studies and tests performed at ASIPP have been effective in clarifying feeder design feasibility and component manufacturability. This paper reports the recent advancements on feeder design, analysis and manufacturing studies. [ABSTRACT FROM PUBLISHER]

The Procurement of the feeder system for The International Thermonuclear Experimental Reactor (ITER) tokamak device started in ASIPP (China) in 2006. The relevant tasks of structural design and analysis have now come to an end, but the R&D activities started in 2009 for critical feeder components and key technology are still ongoing. Following the signature of the Procurement Arrangement between China and the ITER Organization (IO) in 2010, a great number of R&D tasks have been launched with the purpose of testing and verifying the original design, leading to design changes and optimizations. This paper mainly discusses the progress of the feeder design and R&D activities; it also presents some test results. [ABSTRACT FROM PUBLISHER]

As an important method of quality confirmation for the ITER Feeder system insulated components, high voltage test at Paschen condition need to be carried out once after these components are manufactured. The Paschen test is much more efficient for detecting the micro cracks or defects in solid multi-composite high polymer insulation materials than the conventional high voltage test done in atmospheric condition, and almost without any damage for the qualified components. [ABSTRACT FROM PUBLISHER]

Recent design changes to the ITER Poloidal Field (PF) coils and PF conductor layout have been implemented to give a greater operational window for low li (self-inductance) plasmas during burn, extend the operating window for plasmas with currents above 15 MA, and improve the plasma vertical stability control. In addition, the PF and CS (Central Solenoid) operating window has been updated leading to higher currents and peak fields in some of the PF coils. Altogether, this results in higher heat generation in the PF conductors due to AC losses. To confirm that the range of heat loads is acceptable for the PF conductors, a time-dependent thermo-hydraulic analysis of the PF coils has been performed with a model based on the GANDALF code. The deposited heat due to AC losses in the conductors, thermal radiation, thermal conduction, and nuclear heating are given as input data. The results show a moderate impact on minimum temperature margin of the PF conductors during the baseline scenario, stemming from the design modifications. [ABSTRACT FROM AUTHOR]

The ITER Organization (10) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly developing the 68 kA current leads using High Temperature Superconductors for the superconducting Toroidal Field (TF) magnet system of the International Thermonuclear Experimental Reactor, ITER. The proposed design consists of a conventional helium cooled heat exchanger operating between 65 K and 320 K and an HTS module covering the low temperature end using Bi-2223 tapes. The first HTS current lead prototypes will be tested in the EAST tokamak facility at ASIPP. This paper discusses the design of the first, 68 kA, TF HTS lead, which is the main emphasis of this first stage of the ITER current lead development program. [ABSTRACT FROM AUTHOR]

The 2013 MEM workshop is the sixth in the series and proved to be one of the largest, with over 60 participants and over 40 presentations. Only a small selection of these presentations have progressed from PowerPoint to a reviewed paper accepted for publication in Superconductor Science and Technology (SUST). However, the dynamics of a good workshop should not be judged solely by the output of high quality papers. The interplay between electromagnetic and mechanical effects in superconductors is proving to be one of the critical issues in understanding the factors that dominate the practical application of not only the more traditional low temperature superconductors (LTS) but also, perhaps surprisingly, the high temperature superconductors (HTS) materials too. The lessons learned in the design and the diagnostic techniques which have been developed for LTS now overlap into the HTS field. The MEM13 workshop provided an opportunity for informal discussion across these fields, to the benefit of all. The topics of the workshop were selected to allow, as far as practically possible, co-presentation of LTS and HTS materials, as follows: Electro-mechanical effects on practical conductor performance. Analysis of electro-mechanical effects. Diagnosis of electro-mechanical effects. Strain dependence. Simulations and standardized tests for electro-mechanical effects. Microstructure, properties and sensitivity. Coated conductors. Manufacture of LTS and HTS conductors. The papers were wide ranging, covering basic manufacturing issues through to sophisticated testing and advanced diagnosis tools. The MEM13 workshop provided many papers based on the ITER construction, and diagnoses using the large quantity of data available from the industrial scale fabrication and testing of ITER conductors, as well as papers on the frontiers of HTS industrial technologies. It was clear that the huge industrial fabrication of Nb3Sn conductors for ITER, and the associated test programmes, have driven the development of modelling and diagnostic tools to an extent that is exceptional. Assessment of these data, and the application of the conclusions, is only just beginning. We were also fortunate in having industrial contributions from leaders in the various HTS fields, covering production, production issues and critical performance problems. The papers selected for publication are in the more complete areas of development, where sufficient definitive conclusions were available to satisfy the publication requirements of the journal SUST. In the workshop environment, there were plenty of dynamic and lively discussions around the uncertainties. I expect that such discussions will sow the seeds of future developments, appearing in future MEM workshops. Acknowledgments Finally I would like to acknowledge the support and drive of the session moderators in the success of MEM13: A Devred, A Nijhuis, Y Yang, D Larbalestier, K Osamura and A Ballarino. Without them, the workshop and the discussions would have been far less lively and far less structured. [ABSTRACT FROM AUTHOR]

Abstract: The ITER Magnet system consists of 4 main coils sub-systems, i.e. 18 toroidal field coils (TFC), a central solenoid (CS), 6 poloidal field coils (PF) and 3 sets of correction coils (CC). The ITER fusion project has selected the stainless steel 316LN as main material for the magnet structure. The CC contribute to reducing the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. During plasma operation, a large number of loading condition scenarios have been considered and structural analysis performed on key items like Cable-In-Conduit Conductor and the coil case. The results obtained are used for both static and fatigue structural assessment defining the present baseline design. For the construction of the structural cases, welding techniques such as GTAW (Gas Tungsten Arc Welding) and techniques resulting in low distortion and shrinkage like EBW (Electron Beam Welding) or Laser Beam Welding (LBW) with filler metal wire have been selected. Those methods are considered for future qualifications to guarantee proper weld parameters and specified weld properties. In order to determine the strength and fracture toughness of 316LN stainless steel welds with respect to design criteria, some mechanical tests have been carried out at 7K (or 77K), and room temperature. [Copyright &y& Elsevier]

Abstract: The inboard shielding of a fusion reactor can be a crucial issue due to the limited space available in a tokamak configuration. It is necessary to assess the inboard shielding capability of DEMO for its initial design. In this paper, 1D and 3D neutronics models were developed based on a reference design of the Chinese Fusion Engineering Testing Reactor (CFETR). The neutron wall load (NWL) is in the range of 1.5–3MW/m2 and the inboard shielding thickness is constrained within 40–70cm in order to achieve the tritium self-sufficiency of the reactor. Referring to the detailed design of the ITER Toroidal Field Coils (TFCs) and using radiation hardening technology developed for ITER, the inboard blanket shielding capability and nuclear responses of the TFC are investigated for both FLiBe and Li4SiO4 breeding blanket concepts. The impact of the gaps on shielding performance is discussed. Some suggestions on improving the inboard shielding performance for DEMO are also proposed. [Copyright &y& Elsevier]