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The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m −2 with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet temperatures equal to 285 °C/325 °C respectively and based on geometric design parameters such as channel pitch, diameter of pipes and thicknesses. It became clear through the optimization that the advantages of a waved FW are diminished. Given the manufacturing issues of that concept, the waved FW was therefore not pursued further. Even if the optimization study shows that the maximum heat load could in principle be as high as 2.53 MW m −2 , it is reduced to 1.57 MW m −2 when additional constraints are introduced in order not to affect corrosion, manufacturability and Tritium Breeding Ratio in normal condition such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total FW thickness ≤22 mm. However it is important to note that the FW channels currently fulfill additional functions and are therefore not optimized “at all cost” regarding heat load capacity and the paper points out some recommendations against missing assumptions. [ABSTRACT FROM AUTHOR]

In order to fully validate “ITER-like” actively water cooled tungsten plasma facing units, addressing the issues of long plasma discharges, an axisymmetric divertor structure has been studied and manufactured for the implementation in the WEST (W-Tungsten Environment in Steady state Tokamak) tokamak platform. This assembly, called divertor structure and coils (4 m diameter, 20 t), is composed of two stainless steel casings containing an actively water cooled (up to 180 °C, 4 MPa) copper winding pack designed for a conductor current in the range of 12.5kA (up to 1000s). It must sustain harsh environmental conditions in terms of ultra-high vacuum, high temperatures and electrodynamic loads. One major difficulty is the assembly by induction brazing of individual bended conductor sectors inside the vacuum vessel and the consecutive sealing of the casings by TIG welding. Therefor development activities have been carried out on a scale one dummy coil, such as brazing, assembly, thermal cycling and electrical insulation tests (5 kV ground voltage). Whereas the brazing assembly technics and the conductor installation were validated without major difficulties, different technical solutions for the electrical insulation had to be tested. The chosen solution is a resin epoxy impregnated fiber glass fabric layered around the conductors followed by a polymerization procedure. In parallel the manufacturing of divertor structure components started in the second half of 2013 with a total delivery at the end of 2015. The paper will illustrate the technical developments which have been performed in order to fully validate the design. It concerns mainly the dummy coil and the complex conductor installation procedure assisted by virtual reality tools. The manufacturing methods proposed by industry in order to fulfil the technical requirements will be also addressed. Finally the processes and associated tools used in order to implement this large component inside the WEST vacuum vessel will be detailed. [ABSTRACT FROM AUTHOR]

The ITER central interlock system (CIS) architecture is composed of four categories of hardware: the fast architecture, slow programmable logic controller (PLC)-based architecture, hardwired architecture, and servers. The CIS fast architecture receives interlock events from various local plant systems of ITER and communicates the corresponding actions to any other local plant systems in order to avoid or mitigate the damage to the machine. Such functions require a reaction time in the range of 1–10 ms, which is faster than the capability of a PLC. The CIS fast architecture consists of a module called the plasma protection module (PPM), which is mainly in charge of interlock functions related to the plasma. In addition to satisfying the time performance requirements, the PPM complies with the CIS reliability, availability, and integrity requirements (a probability of failure per hour of less than 10E-7 and a failsafe solution). In this paper, we explain the engineering design of our approach from a technical perspective. A COTS FPGA in a redundant configuration solution, which uses serial communication with the local plant systems, is considered. [ABSTRACT FROM AUTHOR]

The paper presents an attempt to proceed to a general concept of software environment for fast and consistent multi-task simulation of EM transients (engineering simulator for tokamak applications). As an example, the ITER tokamak is taken to introduce a computational technique. The strategy exploits parallel processing with optimized simulation algorithms based on using of influence functions and superposition principle to take full advantage of parallelism. The software has been tested on a multi-core supercomputer. The results were compared with data obtained in TYPHOON computations. A discrepancy was found to be below 0.4%. The computation cost for the simulator is proportional to the number of observation points. An average computation time with the simulator is found to be by hundreds times less than the time required to solve numerically a relevant system of differential equations for known software tools. [ABSTRACT FROM AUTHOR]

Toroidal field (TF) superconducting magnets are the key components of the China Fusion Engineering Test Reactor (CFETR). While the TF magnets operate in a steady-state, they are located in a time-varying poloidal field environment, which may lead to AC losses. In this paper, the AC losses of a CFETR TF magnet during excitation and plasma disruption were analyzed. According to the experimental data, the relationship between the critical current density and the magnetic field (J c –B) was obtained, allowing the hysteresis loss of a TF magnet to be calculated. The results show that the hysteresis loss during excitation is high, at about 1.26 MJ. The coupling loss of a TF magnet can be evaluated when the time constant of a conductor is obtained theoretically. The AC losses of TF magnet joints were analyzed by the same method, which showed that they are much less than those of a TF magnet. [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]

This paper describes the methodology that has been developed and applied to derive the principal geometry of the main DEMO tokamak systems, in particular the radial and vertical cross section based on the systems code output parameters, while exact parameters are described elsewhere [1] . This procedure reviews the analysis of the radial and vertical build provided by the system code to verify critical integration interfaces, e.g. missing or too large gaps and/or insufficient thickness of components, and updates these dimensions based on results of more detailed analyses (e.g. neutronics, plasma scenario modelling, etc.) that were carried out outside of the system code in the past years. As well as providing a 3D configuration model of the DEMO tokamak for integrated engineering analysis, the results can also be used to refine the systems code model. This method, subject to continuous refinement, controls the derivation of the main machine parameters and ensures their coherence vis-à-vis a number of agreed controlled physics and engineering assumptions. [ABSTRACT FROM AUTHOR]

To improve the plasma vertical stabilization during tokamak operation, partial pure copper layer could be considered on the inner shell of vacuum vessel which is large complex component made of stainless steel. The copper cladding has been a great challenge due to the complicated configuration and difficulty of dissimilar metal joining. From an engineering point of view, there are several candidates for the copper cladding such as cold spray, overlay welding and explosive welding. In this paper, fabrication feasibility study for copper cladding has been carried out by using cold spray method, and a set of full scale mock-up of in-vessel component has been conducted to demonstrate applicability of copper cold spray and to verify copper layer. The results of copper cold spray on test specimens and mock-up showed good properties and no defects on the thick copper layer. Optimum conditions of copper cold spray have been developed based on these results. Accordingly, it can be concluded that cold spray is a feasible and practical method to achieve thick copper cladding on tokamak component. [ABSTRACT FROM AUTHOR]

The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses results, and prove of the soundness of the design are presented. [ABSTRACT FROM AUTHOR]

Due to this position close to the plasma, the port plug structure and the diagnostic first wall (DFW) contain water to allow cooling during operation and for heating during bake-out. To remove the heat coming from the plasma due to radiation and neutrons, the pressure inside these structures should be up to 44 bars. On the other hand, the dominant load expected to drive the design of these structures is of electromagnetic origin during the plasma disruption. Description of the loads acting on DFWs and generic port plug structures and the significance of the load due to the water pressure, with implications on the design and inspection, are discussed in this paper. [ABSTRACT FROM AUTHOR]

The test facility (TF) is one of the three major facilities of the International Fusion Material Irradiation Facility (IFMIF). Engineering designs of TF main systems and key components have been initiated and developed in the IFMIF EVEDA (Engineering Validation and Engineering Design Activities) phase since 2007. The related work covers the designs of a test cell which is the meeting point of the TF and accelerator facility and lithium facility, a series of test modules for experiments under different irradiation conditions, an access cell to accommodate remote handling systems, four test module handling cells for test module processing and assembling, and test facility ancillary systems for engineering support on energy, media, and control infrastructure. This paper summarizes the principle functions, brief specifications, and the current design status of the above mentioned IFMIF TF systems and key components. [ABSTRACT FROM AUTHOR]

[Display omitted] • Industrial production of ferritic oxide dispersion strengthened steel plates. • Dispersion strengthened first wall; mockup fabrication by scalable processes. • High heat flux fatigue test of a helium cooled breeder blanket first wall. • Increased operating temperature does not lead to damage in the first wall. • Possible strategy for mitigation of irradiation damage in plasma-facing components. A major part in the EUROfusion materials research program is dedicated to characterize and quantify nuclear fusion specific neutron damage in structural materials. While the majority of irradiation data gives a relatively clear view on the displacement damage, the effect of transmutation – i.e. especially hydrogen and helium production in steels – is not yet explored very well. However, few available results indicate that EUROFER-type steels will reach their operating limit as soon as the formation of helium bubbles reaches a critical amount or size. At that point, the material would fail due to embrittlement at the considered load. This paper presents a strategy for the mitigation of the before-mentioned problem using the following facts: • the neutron dose and related transmutation rate decreases quickly inside the first wall, that is, only a plasma-near area is extremely loaded • nanostructured oxide dispersion strengthened (ODS) steels may have an enormous trapping effect on helium and hydrogen, which would suppress the formation of large helium bubbles • compared to conventional steels, ODS steels show improved irradiation tensile ductility and creep strength In summary, producing the plasma facing, highly neutron and heat loaded part of blankets by an ODS steel, while using EUROFER97 for everything else, would allow a higher heat flux as well as a longer operating period. Consequently, we (1) developed and produced 14 % Cr ferritic ODS steel plates. (2) We fabricated a mockup with 5 cooling channels and a plated first wall of ODS steel, using the same production processes as for a real component. And finally, (3) we performed high heat flux tests in the HELOKA facility (Helium Loop Karlsruhe at KIT) applying short and up to 2 h long pulses, in which the operating temperature limit for EUROFER97 (i.e., 550 °C) was finally exceeded by 100 K. Thereafter, microstructure and defect analyses did not reveal defects or recognizable damage. Only a heat affected zone in the EUROFER/ODS steel interface could be detected. This demonstrates that the use of ODS steel could make a decisive difference in the future design and performance of breeding blankets. [ABSTRACT FROM AUTHOR]

• This paper evaluated the capability of electron cyclotron emission (ECE) diagnostic in the Chinese Fusion Engineering Testing Reactor (CFETR), by simulated ECE spectra using the code SPECE. • The results indicate that measurement of X-mode and O-mode ECE spectra combined are capable of providing the information of local electron temperature with fairly good spatial resolution and adequate radial range. • Followed with the simulation results, the conceptual design of the front-end, transmission line, instrumentation, and the calibration strategy are discussed briefly. To evaluate the capability of electron cyclotron emission (ECE) diagnostic in the Chinese Fusion Engineering Testing Reactor (CFETR), ECE spectra have been simulated using the code SPECE. The results indicate that measurements of the 2nd harmonic X-mode and the 1st harmonic O-mode ECE spectra combined are capable of providing the information of local electron temperature with fairly good spatial resolution and adequate radial range covering from the plasma core to the edge. Followed with the simulation results, the conceptual design of the front-end, transmission line, instrumentation, and the calibration strategy are discussed briefly. Base on the state-of-the-art techniques, no foreseen technical obstacles exist. However, it is worthy to study the relativistic down-shift effect in the high temperature plasmas in order to obtain the position of the emission layer with required precision in a timely manner. [ABSTRACT FROM AUTHOR]