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The utilization of composites layered on an additive manufactured substrate, for the production of coil supports for modular coils in small or middle size experimental stellarators is assessed. The focus of the study is a monolithic coil support comprising the coils of a half-period of a stellarator, somewhat similar to the ones in UST_2 and ARIES-CS stellarators. However, the concept may be applicable to quasi-monolithic coil supports (coil forms of NCSX type) or individual coil casings (W7-X type). Coil supports for stellarators require high precision, stiffness and strength for large contorted parts. Traditionally, monolithic coil supports are produced by steel casting/forging and final machining. This production method and material gives accurate, stiff and strong coil supports, but the method may be expensive for monolithic supports due to the geometrical complexity and required accuracy of the structure. In relation to those matters, this work investigates whether a monolithic coil support comprising an additive manufactured resin substrate, which is externally (outward from the coils) surrounded with a thick layer of fibre-reinforced resin, may achieve enough strength and stiffness under middle/high magnetic fields. Finite element calculations are produced to obtain the direction of the principal stresses and their values in compression and tension at different areas of the monolithic support, which is relevant for anisotropic materials. The feasibility of directional application of (carbon) fibres on the winding surface of the stellarator outward from the coils is experimentally tested on a scaled-down additively manufactured prototype of a monolithic support. The strength and stiffness of the composite structure appears sufficient for common magnetic fields in experimental stellarators, and the 3D-composite design and manufacturing was technically feasible.

The paper studies the dimensional accuracy of additively manufactured (AM) coil winding structures for small experimental stellarators fabricated under industrial non-laboratory conditions. Insufficient accuracy is one of the main issues hindering the use of additive manufacturing (AM) for coil windings structures (coil casings and coil forms or frames) for stellarators. The dimensional accuracy of one complex modular coil frame and four planar coil casings is studied for different AM techniques and materials, in particular Selective Laser Sintering (SLS) in polyamide, Stereolithography (SLA) in resin, Fused Deposition Modelling (FDM) in ABS and PolyJet in resin. The measurements are performed by a Mitutoyo Coordinate Measuring Machine. The AM parts are hollow for subsequent internal casting with (fibre-reinforced) resin for strength and performance enhancement, method named 3Dformwork. The paper reports the features of the sample parts, the metrology methodology utilised, the performed measurements and the resulting dimensional deviations for each AM technique. Professional AM in FDM showed deviation between nominal dimensions and measurements of ±0.1% (one sigma, 68% of measurements) and PolyJet ±0.15%. Personal web-based AM in SLA exhibited deviation ±0.3% and SLS in polyamide ±0.4% (one sigma). The PolyJet part showed dimensional instability under harsh environment and would require immediate 3Dformwork. The FDM part presented the lower cost for the particular case study among the professional AM. Thus, high quality PolyJet and FDM additive manufacturing in plastics are at the verge of achieving the requirement of 0.1% minimum accuracy for stellarator coil windings. Therefore, AM may contribute to the fabrication of accurate winding structures for stellarators and other accurate components in fusion.

The ITER equatorial visible and infrared Wide Angle Viewing System (WAVS) provides surface temperature measurements of plasma facing components by infrared (IR) thermography being one of its main roles to protect them from damage. It also images the plasma emission in the visible range. The diagnostic, which comprises 15 lines of sight distributed in four equatorial ports (3, 9, 12 and 17), will contribute to Machine protection, Basic control and Physics analysis in ITER. The design of the diagnostic in the equatorial port 12 (EP12) - comprising three lines of sight - is currently being developed within a Framework Partnership Agreement through a Specific Grant granted by F4E to the Consortium formed by CEA, Ciemat and Bertin Technologies. EP12 has to be operative for the ITER first plasma. The overview of the WAVS opto-mechanical design for the ex-vessel components in EP12 will be presented in the paper. The ex-vessel optics are composed of mirrors and refractive elements which relay the exit image of the port plug at the vacuum window up to the cameras placed inside a shielded cabinet in the Port Cell. The image relay through the Interspace, Bioshield and Port Cell has to cover a length of almost 10 m in order to reach the detectors. The refractive elements are arranged in sub-assemblies (optical modules) mechanically mounted in barrels and next assembled on a support structure. The assembly of these modules on site will be also presented in the paper. Furthermore, the Optical Hinge which is the first component of the ex-vessel optical chain (formed by two folding mirrors) aims to compensate the differential motion between the vacuum vessel and the building, in normal operation and baking. The mechanical assembly and the operation of this component will be also described in the paper.

Stellarator advancement is hindered, among others, by the requirement of geometric complexity at high accuracy and the still scarce universities and research centres following the stellarator line. In this framework, the objectives of the small UST_1 stellarator development were to: (i) explore and test the performance of one possible accurate construction method for stellarators (ii) encourage universities and small fusion research centres to build simple and economical stellarators (iii) educative purpose. Therefore, UST_1 was properly designed to be easily built by a milling machine working on toroidal coordinates, being the winding surface circular poloidally and toroidally. The coil frame is a sole monolithic toroidal thick surface equipped with grooves mechanised by the toroidal milling machine. Only one double pancake is wound in each groove so as to compress the conductor on the laterals of the groove in order to speed up and simplify the winding process. The physics design, the conceptual engineering design and the construction process of UST_1 is presented. The toroidal milling machine is described. The e-beam field line mapping experiments carried out to validate the resulting magnetic configuration are reported. The developed construction method has been proved for the small UST_1 stellarator. Small stellarators are valuable for quick tests of diagnostics, educative purposes, assessment of new confinement concepts, turbulence studies and other applications.

The geometric complexity and high accuracy simultaneously required in magnetic fusion devices, particularly stellarators and tokamaks, hampers the production of fusion components and devices. Rapid manufacturing construction methods, particularly enhanced for fusion, may contribute to a faster cycle and lower cost production of certain components for tokamaks and stellarators. Casting, cutting, forming, welding and mechanising are conventional production techniques for major fusion components, i.e. coil casings, coil frames, vacuum vessels and blankets. Synergies may emerge by combination of additive manufacturing (3D printing) with conventional manufacturing methods. 3D printing combined with resin moulding is tested by construction of the coil frame and the vacuum vessel of a small stellarator, UST_2. Satisfactory coil frames have been obtained by moulding acrylic resin in a special 3D printed polyamide hollow three-dimensional structure. The conceptual engineering design, construction process and validation of the components are described. The presented manufacturing method might contribute to advance the future 3D printing of larger metallic components for fusion.

The Stellarator of Costa Rica 1 (SCR-1), located at the Instituto Tecnologico de Costa Rica, is the first Stellarator of Latin America [1]. In Stellarators, vacuum magnetic flux surfaces are produced only by external coils [2], making respective property measurements possible. Flux surfaces allow the magnetic confinement of plasma, making their characterization important to verify the device’s construction. The aim of this work is to present measurements on the flux surfaces in the SCR-1. Mapping of the magnetic surfaces produced in the SCR-1 implied the use of a transparent fluorescent screen covered with ZnO:Zn phosphor which allows the detection of an electron beam, emitted from an electron gun inside the vacuum vessel, that travels around the device following the magnetic field lines. Three highly transparent screens were used: a fixed stainless-steel mesh, an oscillating rod, and an in-house elliptical design called “cuchillo de mantequilla” (butter knife); and results were compared between them. Using the three different methods in the toroidal position 0°, three different nested surfaces were measured. Finally, results were compared with the Poincare maps from BS-SOLCTRA (Biot-Savart Solver for Compute and Trace Magnetic Fields) and EXTENDER_P (formerly Poincare) codes [3], [4]. The obtained results are consistent with expectations, verifying that the manufacturing process of the coils was handled as designed.

Geometric complexity of stellarators hampers a straightforward production of conceived optimised magnetic configurations. Integration of the engineering design with new fabrication methods may reduce the production cost and accelerate the production process. A fast cycle production of experimental fusion devices also might result in a faster advance in fusion plasma science. Several different stellarators could be used to test configurations for improved turbulent transport or to validate new divertor configurations. In this framework, and based on the results from the previously built UST_1 stellarator, the present work try to study and validate the feasibility of 3D printing methods (additive manufacturing) for small experimental stellarators. The paper summarises the engineering development, fabrication and validation of a coil frame test sector for the UST_2 stellarator. The definition of the Last Closed Flux Surface and winding surface for the test sector is based on an optimised quasi-isodynamic poloidal stellarator, modified for enhanced in-vessel remote handling manipulation and wide space for divertors. A Filled-sparse coil frame concept is developed to still keep low the cost in spite of the present expensive 3D printing materials and printers.

Physics of stellarators has been optimized up to excellence levels. In contrast, optimization and innovation in the engineering and construction of stellarators is still comparatively modest. Particularly for stellarators but also for tokamaks, geometric complexity, small dimensional tolerances and large size currently imply relatively high fabrication costs, delaying the evolution of the stellarator research line and also hindering tokamaks. In this framework, the small low-cost modular stellarator UST_1, built and operated in a private laboratory, may contribute to the background of cost-reasonable construction methods for small fusion devices, particularly stellarators. The paper summarizes the engineering development and fabrication of UST_1 coils and outlines possible engineering R&D lines for the enhancement of the UST_1 construction methods. UST_1 consists of 12 resistive modular coils wound in grooves mechanised on a single monolithic toroidal frame surrounding the vacuum vessel. A special milling machine working on toroidal coordinates was developed, tested and built in order to accurately mechanise the grooves in the monolithic frame.

EFDA finances a training programme called Goal Oriented Training Programme for Remote Handling (GOT RH), whose goal is to train engineers in Remote Handling for ITER. As part of this training programme, the conceptual design of the mechanical interface between Upper Port Plug (UPP) and Cask and Plug Remote Handling System (CPRHS) as well as the conceptual design of the needed tools for UPP Remote Handling is carried out. The paper presents the conceptual design of the UPP/Gripping Tool Interface. This includes the conceptual design of the gripping tool for introducing/removing the UPP in/from the ITER port and the mechanical features on both sides of the UPP/Gripping Tool Interface (e.g. alignment features, mechanical connectors, fasteners). In order to develop the design of the interface between UPP and CPRHS it is necessary to first identify the functional requirements of the Transfer Cask System (TCS) and the CPRHS, such as required degrees of freedom (DoF), required performances of system, geometrical constraints, loading conditions, alignment requirements, RAMI requirements. These requirements are the input data for the design of the interface between UPP and gripping tool and some of them are also described in the paper.

One of the most technically challenging activities of the IFMIF facility is the maintenance and the refurbishment of its components, and among these the target system appears to be critical since it is located in the most severe region of neutron irradiation (60 dpa/fpy). Two different target assembly systems have been developed: the first is known as integral target while the second one is based on the so called replaceable backplate bayonet concept. The present remote handling (RH) procedures developed for the refurbishment of the removable backplate foresee the removal of all the components from the upper part of the test cell. This operation has a strong impact on the intervention time for the backplate refurbishment which has to be repeated at least every year. Consequently the need to update the RH strategy for the refurbishment of this component becomes a precondition in order to fulfill the stringent requirement to enhance the duty cycle of IFMIF plant. In this paper two potential approaches are presented: the first relies on the possibility to perform all the refurbishment operations in situ in the test cell cavern, whilst the second one foresees to perform these operations off-line in a hot cell. Advantages and disadvantages of these approaches together with the RH requirements for the refurbishment operations of the backplate bayonet concepts are also reported.

The International Fusion Materials Irradiation Facility (IFMIF) is a future neutron source based on the D-Li stripping reaction, planned to test candidate fusion materials at relevant fusion irradiation conditions. During the design of IFMIF special attention was paid to the structural materials for the blanket and first wall, because they will be exposed to the most severe irradiation conditions in a fusion reactor. Also the irradiation of candidate materials for solid breeder blankets is planned in the IFMIF reference design. This paper focuses on the assessment of the suitability of IFMIF irradiation conditions for testing functional materials to be used in liquid blankets and diagnostics systems, since they are been also considered within IFMIF objectives. The study has been based on the analysis and comparison of the main expected irradiation parameters in IFMIF and DEMO reactor.

Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al2O3, SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

A coherent design of the remote handling system with the design of the components to be manipulated is vital for reliable, safe and fast maintenance, having a decisive impact on availability, occupational exposures and operational cost of the facility. Highly activated components in the IFMIF facility are found at the Test Cell, a shielded pit where the samples are accurately located. The remote handling system for the Test Cell reference design was outlined in some past IFMIF studies. Currently a new preliminary design of the Test Cell in the IFMIF facility is being developed, introducing important modifications with respect to the reference one. This recent design separates the previous Vertical Test Assemblies in three functional components: Test Modules, shielding plugs and conduits. Therefore, it is necessary to adapt the previous design of the remote handling system to the new maintenance procedures and requirements. This paper summarises such modifications of the remote handling system, in particular the assessment of the feasibility of a modified commercial multirope crane for the handling of the weighty shielding plugs for the new Test Cell and a quasi-commercial grapple for the handling of the new Test Modules.

A diagnostic based on a set of fission ionization chambers is proposed for the IFMIF Test Cell in order to achieve on-line control of the irradiation parameters. These chambers are widely used in fission reactors; however the IFMIF environment (neutron spectrum, high neutron fluxes, gamma doses, reduced space, etc. …) requires a careful design of the whole diagnostic. This work shows the current status of the diagnostic development, which is being performed in the framework of the Broader Approach activities. The definition of the diagnostic strategy is addressed to establish the main specifications and requirements of the IFMIF fission chambers and their deployment. The map that the signal of a set of fission chambers will supply has been assessed by taking into account their sensitivity, their positioning inside the Test Cell and the IFMIF neutron flux. The results show that this experimental system will be capable of measuring and characterizing the high neutron fluxes expected for IFMIF.

The engineering design of fission chambers as on-line radiation detectors for IFMIF is being performed in the framework of the IFMIF-EVEDA works. In this paper the results of the experiments performed in the BR2 reactor during the phase-2 of the foreseen validation activities are addressed. Two detectors have been tested in a mixed neutron-gamma field with high neutron fluence and gamma absorbed dose rates, comparable with the expected values in the HFTM in IFMIF. Since the neutron spectra in all BR2 channels are dominated by the thermal neutron component, the detectors have been surrounded by a cylindrical gadolinium screen to cut the thermal neutron component, in order to get a more representative test for IFMIF conditions. The integrated gamma absorbed dose was about 4 × 1010 Gy and the fast neutron fluence (E > 0.1 MeV) 4 × 1020 n/cm2. The fission chambers were calibrated in three BR2 channels with different neutron-to-gamma ratio, and the long-term evolution of the signals was studied and compared with theoretical calculations.

Tritium release experiments using different breeding material candidates are planned for the medium flux region of the IFMIF Test Cell. Nowadays, only ceramic breeder materials have been suggested to be tested in the Tritium Release Module located in the Medium Flux Test Module of IFMIF. Liquid breeder blankets are very promising and for that reason, several concepts will be tested in ITER. One of the main problems concerning the liquid blankets is the permeation of the generated tritium in the breeder throughout the walls. Since tritium permeation is highly influenced by irradiation conditions, IFMIF is a suitable scenario to perform tritium permeation related experiments. In this paper, a preliminary design of a tritium permeation experiment for the Medium Flux Test Module of IFMIF is proposed, in order to contribute to the progress of the liquid breeder blanket concept validation. The conceptual design of the capsule in which the experiment will be performed is carried out, taking into consideration the experiment necessities and its implementation in the Tritium Release Module. In addition to this, some thermal hydraulic calculations have been performed to evaluate the thermal behaviour of the irradiation capsule.

High accuracy, geometric complexity and thus high cost of stellarators tend to hinder the advance of stellarator research. Nowadays, new manufacturing methods might be developed for the production of small and middle-size stellarators. The methods should demonstrate advantages with respect common fabrication methods, like casting, cutting, forging and welding, for the construction of advanced highly convoluted modular stellarators. UST_2 is a small modular three period quasi-isodynamic stellarator of major radius 0.26 m and plasma volume 10 litres being currently built to validate additive manufacturing (3D printing) for stellarator construction. The modular coils are wound in grooves defined on six 3D printed half period frames designed as light truss structures filled by a strong filler. A geometrically simple assembling configuration has been concocted for UST_2 so as to try to lower the cost of the device while keeping the positioning accuracy of the different elements. The paper summarizes the construction and assembling concepts developed, the devised positioning methodology, the design of the coil frames and positioning elements and, an initial validation of the assembling of the components.

This paper describes briefly the design and construction of a small modular stellarator for magnetic confinement of plasma, called Stellarator of Costa Rica 1, or SCR-1; developed by the Plasma Physics Group of the Instituto Tecnologico de Costa Rica, PlasmaTEC. The SCR-1 is based on the small Spanish stellarator UST_1, created by the engineer Vicente Queral. The SCR-1 will employ stainless steel torus-shaped vacuum vessel with a major radius of 460.33 mm and a cross section radius of 110.25mm. A typical SCR-1 plasma will have an average radius 42.2 mm and a volume of 8 liters (0.01 m3), and an aspect ratio of 5.7. The magnetic resonant field will be 0.0878 T, and a period of 2 (m=2) with a rotational transform of 0.3. The magnetic field will be provided by 12 modular coils, with 8 turns each, with an electrical current of 8704 A per coil (1088 A per turn of each coil). This current will be fed by a bank of cell batteries. The plasma will be heated by ECRH with magnetrons of a total power of 5kW, in the first harmonic at 2.45GHz. The expected electron temperature and density are 15 eV and 1017 m−3 respectively with an estimated confinement time of 7.30 x 10−4 ms. The initial diagnostics on the SCR-1 will consist of a Langmuir probe, a heterodyne microwave interferometer, and a field mapping system. The first plasma of the SCR-1 is expected at the end of 2011.

The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept, the re-location of the quench tank outside the test cell (TC) with a reduction of tritium inventory and a simplification on its replacement in case of failure, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction, and the water cooling of the liner and biological shielding of the TC, enhancing the efficiency and economy of the related sub-systems. In addition, the maintenance strategy has been modified to allow a shorter yearly stop of the irradiation operations and a more careful management of the irradiated samples. The design of the IFMIF plant is intimately linked with the EVA phase carried out since the entry into force of IFMIF/EVEDA in June 2007. These last activities and their on-going accomplishment have been thoroughly described elsewhere (Knaster J et al [19]), which, combined with the present paper, allows a clear understanding of the maturity of the European-Japanese international efforts. This released IFMIF Intermediate Engineering Design Report (IIEDR), which could be complemented if required concurrently with the outcome of the on-going EVA, will allow decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community. © 2015 IAEA, Vienna.