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The corrosion induced by alkali metals in steels has been the subject of long decades of intense studies under both nuclear fission and fusion research programs. Li or its eutectic Pb-17Li is the liquid metal coolant choice for fusion blankets due to the tritium breeder capability of Li. Non-metal impurities enhance corrosion, but only N becomes potentially a problem given its high solubility in liquid Li and the depletion of Cr through ternary nitrides Li-Cr-N. The low solubility of C and O allow its cold trapping to values <10 wppm, however N can only be hot trapped demanding temperatures typically of 873 K. The inherent difficulties of experimentation on physicochemical kinetics related with alkali metals lead to a confusing divergence of results available in the literature; however, the understanding of the corrosion phenomena of RAFM steels exposed to flowing Li up to 873 K is mature. Next decade, 14 MeV neutrons will be available for fusion materials testing through Li(d,n) nuclear reactions. In such a facility, a concave RAFM steel backplate will be channelling 523 K flowing Li in the region where the 40 MeV deuteron beam will be impacting. If RAFM steels are considered, two main concurrent mechanisms will take place: a) mass transport of alloying elementsalong the loop and b) depletion of Cr through formation of Li 9 CrN 5 . Fortunately, the mass transport phenomena of Cr within the ΔT = 350 K in the loop is limited due to the poor solubility of Cr in liquid Li (0.21 wppm at 873 K). In turn, at 523 K Li the activity of N to form the ternary compound is negligible. However, the high solubility of Ni in Li (2144 wppm at 873 K), suggests the presence of mass transport phenomena of Ni from the stainless steel piping; unfortunately, the physicochemical kinetics are not fully understood. Lifus 6, in operation in Brasimone (ENEA) since the end 2015, will close in a definitive manner remaining open questions. [ABSTRACT FROM AUTHOR]

Abstract: The International Fusion Materials Irradiation Facility (IFMIF) Engineering Design and Engineering Validation Activities (EVEDA) are being developed in a joint project in the framework of the Broader Approach (BA) Agreement between EU and Japan. This project has now entered into a crucial phase as the engineering design of IFMIF is now being formulated in a series of 3 subsequent phases for delivering an Interim IFMIF Engineering Design Report (IIEDR) by mid of 2013. Content of these phases is explained, including the plant configuration detailing the 5 IFMIF facilities and their systems. Together with the Engineering Design Activities, prototyping sub-projects are pursued in the Engineering Validation Activities which consist of the design, manufacturing and testing of the following prototypical systems: Linear IFMIF Prototype Accelerator (LIPAc), EVEDA Lithium Test Loop (ELTL), and High Flux Test Module (HFTM) with the prototypical helium cooling loop (HELOKA). Highlights are described from recent experiments in the Engineering Validation Activities. [Copyright &y& Elsevier]

Abstract: Error fields in Tokamaks are small departures of the exact axisymmetry of the ideal magnetic field configuration. Their reduction beyond a threshold value by the error field correction coils is essential since sufficiently large static error fields lead to discharge disruption. The error fields are originated not only by coils fabrication and installation alignment tolerances, joints and busbars but also due to the presence of ferromagnetic elements. The start of plasma current flattop with relatively low plasma density is considered as a critical state of the 15MA scenario for the onset of locked modes causing disruptions. A figure of merit, the ‘3-mode’ criterion, based on the lowest error field harmonics [(1,1); (2,1); (3,1)] was chosen to assess the error fields expected in ITER. Analysis performed last years in an independent way by CREATE (EU) and Efremov Inst. (RF) groups has allowed a deep understanding of the error fields induced by all the possible sources. Both groups were successfully benchmarked with the estimation of the error fields induced by a given deformed shape of a TF coil obtaining a perfect match of the results up to 2 orders of magnitude smaller than the defined threshold of B 3-mode/B to <5×10−5. Three different sets of independent variables based on a 3D rigid body movement of the coils have been provided for the 6 CS modules, 18 TF coils and 6 PF coils tolerances, which have allowed a clear understanding of the weight of each of the variables in the induced error fields. The results obtained in 2008 with a realistic set of magnets tolerances concluded that the system of correction coils provides effective suppression of the error fields with margin to correct possible impact of other sources. In particular, it has been shown that superconducting joints, feeders and busbars play a secondary effect; however the radial position of the TF coils and the tilt and radial shift of the CS stack would have a relevant influence. The ensuing recent sets of variables studied aimed at deepening the understanding of the operational limits. The present paper addresses the impact of the fabrication tolerances and installation misalignments of the magnet system; the impact of ferromagnetic inserts and busbars as well as the design of the correction coils will be thoroughly covered elsewhere (Amoskov et al., in press ). [Copyright &y& Elsevier]

Abstract: The toroidal field (TF) coils of ITER will undergo out-of-plane forces caused by the poloidal fields required to confine the plasma. These forces will be supported against overturning moments by links between the coils. In turn, these links consist of the inner intercoil structure (IIC), which is composed of two pairs (placed at the top and bottom part of the inboard leg) of four sets of poloidal shear keys inserted in slots between adjacent coils, and the outer intercoil structure (OIS) formed by four bands of shear panels on the outboard leg. The magnetic forces during energization of ITER would cause at IIC locations at toroidal gap between adjacent TF coils of 0.35mm; during plasma operation this value could reach >1mm causing a loosening of the keys and intensifying stress concentrations. This undesired effect will be suppressed by the application of a centripetal force of 70MN/coil (35MN at both the bottom and top part of the inboard leg of each of the 18 TF coils) that will be provided by two sets of three glass fibre/epoxy composite rings submitted to a toroidal hoop force of 100MN/set. The calculated maximum stress in the rings will occur during the installation phase at room temperature, where the maximum radial elongation (∼25mm) is required, and it reaches 1/5 of the composite presently estimated ultimate stress. The imposed elongation to reach that force and the lower Young''s modulus of the composite compared with that of stainless steel will ease component tolerances and/or settlement effects in the final assembly. The paper describes the evolution in the design of the pre-compression system, from the conceptual phase when two circular cross-sections rings were considered to the present definitive one with three rectangular cross-section rings. [Copyright &y& Elsevier]

The International Fusion Material Irradiation Facility (IFMIF) aims at qualifying and characterizing materials capable to withstand the intense neutron flux originated in the D-T reactions of future fusion reactors by a neutron flux with a broad peak at 14 MeV, capable to provide >20 dpa/fpy on small specimens, also qualified in this engineering validation engineering design activity (EVEDA) phase. Its broad mandate has been successfully achieved, only pending the validation of its accelerator with its conventional facilities. The validation of the IFMIF’s accelerators will be achieved in this on-going phase, until December 2019, with the operation of a deuteron accelerator at 125 mA CW mode and 9 MeV, which is presently under installation and commissioning in Rokkasho (Japan). The target availability of the IFMIF facility, 70%, is one of its main challenges because it demands an extraordinary individual availability of the sub-systems, such as the accelerator, with 87%. The linear IFMIF prototype accelerator (LIPAc) presents a broad spectrum of ancillary equipment to optimize its operational beam time. A description of the nuclear HVAC of IFMIF has already been reported (Pruneri et al., 2016) [1] . The present paper describes the design of the conventional systems of LIPAc, among which we address the electrical power supply, the heating, ventilation, and air conditioning (HVAC), the heat rejection system (HRS), the service water system (SWS), the service gas system (SGS), the cryoplant system (Cryo), and the fire protection system (FPS). [ABSTRACT FROM AUTHOR]

To show the Hertz-Knudsen equation is applicable to the prediction of the evaporation rates of Li and volatile impurities from the Li target surface at the International Fusion Materials Irradiation Facility (IFMIF), we analytically and experimentally studied these rates for a high-speed Li jet via engineering-scale tests performed under IFMIF conditions (vacuum pressure: 10 −3 Pa, Li temperature: 250 °C, jet speed: 7–15 m/s). The rates of evaporation and deposition were analytically predicted using the Hertz–Knudsen equation and the cosine law with known Li impurity concentrations (36 wppm for Na and 10 wppm or less for K). As pointed out in earlier works, Na and K are volatile impurities in Li. The evaporation rates of Na and K from liquid Li at a temperature of 250 °C are non-negligible, even if their concentrations in Li are as small as tens of wppm. To measure the deposition rate of evaporants from the Li jet, we used a quartz crystal microbalance and a neutralization titration method. The measured deposition rate was highly consistent with the deposition rate estimated analytically for weak evaporation. Also, the measured deposition rate was independent of the flow speed in the range of 7–15 m/s. We concluded that combining the Hertz–Knudsen equation and the cosine law is a reliable method for predicting the evaporation and deposition rates of Li and volatile impurities from a Li jet surface at the IFMIF. [ABSTRACT FROM AUTHOR]

The Broader Approach (BA) activities consists of three major projects: the International Fusion Energy Research Center (IFERC) project, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, and the Satellite Tokamak Programme (STP, JT-60SA). These projects have been carried out to obtain basic data for the design of DEMO fusion reactor from 2007. For 8-year activities, the above projects could produce a set of fruitful results for the DEMO reactor. DEMO design activity has been conducted to build a set of DEMO design bases in accordance with a series of discussion between EU and JA. In the DEMO R&D activities, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. Some additional R&D subjects recommended by peer review comments have also been studied successfully in recent years. Regarding the IFMIF/EVEDA project, some main components of the accelerator facility been designed and tested. The validation test using EVEDA Lithium Test Loop (ELTL) was also completed successfully in October 2014. [ABSTRACT FROM AUTHOR]

Three different partners, ENEA, JAEA ed University of Tokyo, have been involved during 2014–2015 in the Round Robin experimentation for the assessment of the soundness of the analitycal procedure for the determination of the Nitrogen impurities contained inside a solid Lithium sample. Two different kinds of Lithium samples, differing by about an order of magnitude in Nitrogen concentration (∼230 wppm; ∼20–30 wppm), have been selected for this cross analysis. The agreement of the achieved results appears very good for what concerns the most concentrated Lithium and indicates each partner’s procedure is appropriate and intrinsecally able to lead to meaningful values, characterized by a relative uncertainty of just few %. The smaller agreement in the case of the less concentrated Lithium anyway points out that particular attention must be paid to reduce as much as possible any source of external contamination and highlights the importance of the proper blank subtraction. [ABSTRACT FROM AUTHOR]

In 2013, the IFMIF, the International Fusion Material Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase, framed by the Broader Approach Agreement between Japan and EURATOM, accomplished in 2013 its mandate to provide the engineering design of the plant on schedule [1] . The IFMIF aims to qualify and characterize materials that are capable of withstanding the intense neutron flux originated in D-T reactions of future fusion reactors due to a neutron flux with a broad peak at 14 MeV, which is able to provide >20 dpa/fpy on small specimens in this EVEDA phase. The successful operation of such a challenging plant demands a careful assessment of the Conventional Facilities (CF), which have adequate redundancies to allow for the target plant availability [2] . The present paper addresses the design proposed in the IFMIF Intermediate Engineering Design Report regarding the CF, particularly the IFMIF's Nuclear and Industrial HVAC design. A preliminary feasibility study, including the initial configuration, calculations and reliability/availability analysis, were performed. The nuclear HVAC design was developed progressively; first, by establishing a conceptual design, starting from the system functional description, followed by the identification of the corresponding interfacing systems and their technical requirements. Once the technical requirements were identified, safety zones were identified based on the radiation classification, frequency dose and parameter of Derivate Air Contamination (DAC). The zone color was determined to match the room's radiation classification. The system design was further developed by defining and creating a Block Diagram with basic and additional information, eventually resulting in a Process Flow Diagram concurrent with the equipment layout definition. Subsequently, we studied and developed the various Piping & instrumentation diagrams (P&ID's), air duct layout and equipment list for different air handling units, air ducting as well as a layout plan of the equipment piping, which was eventually integrated into the 3D model of the building and coordinated with others subsystems of the IFMIF. [ABSTRACT FROM AUTHOR]

The International Fusion Material Irradiation Facility (IFMIF) is an experimental facility conceived for qualifying and characterizing structural materials for nuclear fusion applications. The Engineering Validation and Engineering Design Activity (EVEDA) is a fundamental step towards the final design. It presented two mandates: the Engineering Validation Activities (EVA), still on-going, and the Engineering Design Activities (EDA) accomplished on schedule in June 2013. Five main facilities are identified in IFMIF, among which the Lithium Target Facility constituted a technological challenge overcome thanks to the success of the main validation challenges impacting the design. The design of the liquid Lithium Target Facility at the end of the EDA phase is here detailed. [ABSTRACT FROM AUTHOR]

A liquid-Li free-surface stream flowing at 15 m/s under a high vacuum of 10 −3 Pa is to serve as a beam target for the planned International Fusion Materials Irradiation Facility (IFMIF). The Engineering Validation and Engineering Design Activities (EVEDA) for the IFMIF is implemented under the Broader Approach. As a major activity for the Li target, the EVEDA Li test loop (ELTL) was constructed by the Japan Atomic Energy Agency. The present study demonstrates a stable Li target that satisfies the IFMIF design conditions (250 °C, 15 m/s, 10 −3 Pa) and presents the procedure for operating the Li target which consists of seven stages. We used a laser-based method to verify that the Li target is adequately stable. In addition, we examine cavitation occurring during startup of the Li target, and consequently, determine the appropriate startup pressure. [ABSTRACT FROM AUTHOR]

Abstract: The linear IFMIF prototype accelerator (LIPAc) has been launched within the framework of the Broader Approach Agreement with the objective to validate the low energy part (9MeV) for the IFMIF linacs (40MeV, 125mA of D+ beam in continuous wave). Starting in mid-2007, the project is managed by the two Home Teams (JA-HT and EU-HT) and headed by the Project Team at the Broader Approach site in Rokkasho with the aim to complete the validation activity with the installation and commissioning of the whole LIPAc by June 2017. Following a short description and current status of each accelerator component, this paper describes the integration activities such as the 3D mockup integration and the Interface Management System (IMS) tools developed for use at the Integrated Project Team level. In preparation of the first equipment delivery in Japan at the beginning of 2013, implementation and installation activities of the various components are described, in particular assembly procedures associated with each subsystem. [Copyright &y& Elsevier]

Abstract: The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R&D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R&D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed. [Copyright &y& Elsevier]

Abstract: In the frame of the ITER coils production, one crucial point is the cover plate welding to complete a coil layer, the so called double pancake (DP). The scope of the activity here described is the assessment of the deformation induced by this laser welding through a FE numerical simulation, in order to find the welding sequence that minimize the global distortion. At this concern, the Local–Global approach technique has been adopted using Abaqus Code, in order to be able to compare several long welding paths in acceptable CPU time increasing analysis speed and model definition easiness in spite of partial loss in the results accuracy. The numerical methodology has been tuned and assessed on experimental data from a welding sample in which a 2.5mm deep welding has been performed and the deformation induced by 1 and 2mm deep welds (the ones that will be used in the complete DP) has been extrapolated. These local strains have then been applied on the global models reproducing the middle and side DP, and their deformation has been computed. The rather good fitting between the experimental and numerical results is very promising, and a complete asses of the procedure could be performed in the frame of an experimental–numerical research campaign. [Copyright &y& Elsevier]

Abstract: Fiber reinforced composites impregnated with mixtures of cyanate ester and epoxy resin demonstrated their excellent performance at the ITER design fluence and beyond. Slight modifications of the resin were necessary to extend the pot-life to more than 100h due to the complexity of the coil structure. Two CE blends were supplied by Huntsman, Switzerland, and CTD, USA, according to the ITER TF coil specification, in order to be qualified for ITER. Sets of test samples were produced under exactly the same conditions. The insulation systems consist of a wrapped R-glass/polyimide reinforcement, vacuum impregnated with the cyanate ester/epoxy blends. The mechanical properties were characterized prior to and after irradiation to a fast neutron fluence of 2×1022 m−2 (E >0.1MeV), i.e. twice the ITER design fluence, in tension and interlaminar shear at 77K. In addition, tension–tension fatigue measurements were carried out in order to simulate the pulsed operation of ITER. The results show that the mechanical properties of both materials are hardly affected by neutron irradiation. [Copyright &y& Elsevier]

Abstract: The ITER pre-compression ring system is based on two sets of 3 rings made in fibreglass-epoxy composite and is located on top and bottom of the inner straight leg region. These rings will provide a radial force of about 70MN/coil (35MN at the top and 35MN at the bottom) under cryogenic conditions, pulling the TF coils into contact, reducing the gap between the poloidal shear keys and their keyways during operation and also reducing the toroidal tension in the four outer intercoil structures bands. The paper describes the stress relaxation test performed in ENEA Frascati laboratories on a reduced scale glass fibre-epoxy composite ring. The ring was manufactured with a diameter of 1m (1/5 of the real scale one) by winding S2 glass fibres then impregnated with epoxy. The volumetric glass content was of 69% and no significant presence of defects was verified by an X-ray survey. The test has been carried out following as close as possible the ASTM standard E 328-86. The ring has been inserted into a dedicated hydraulic testing machine consisting of 18 radial pulling actuators and then loaded up to a ring membrane hoop stress of 800MPa. The deformation has been maintained constant during the test by inserting a solid steel disk in the machine centre restraining the machine radial motion. Average hoop stress has been daily monitored measuring the force required to lift the ring just free of the constraints during the test period. The ring performed 125 testing days showing no significant stress relaxation and a small reduction of the ring stiffness that appeared mainly in the first days of the test. [Copyright &y& Elsevier]

Abstract: The first 2 years of the ITER IO has seen substantial progress towards the construction of the magnets, in three main areas. Firstly, the design has been developed under the conflicting constraints to minimise construction costs and to maximise plasma physics performance. Building construction momentum while updating the design to take account of new physics assessments of the coil requirements has been challenging. Secondly, with a stabilising design, it has been possible for the Domestic Agencies to launch the first industrial procurement contracts. And thirdly, critical R&D to confirm the performance of the Nb3Sn cable in conduit design is proceeding successfully. The design consolidation has been accompanied by design reviews involving the international community. The reviews conducted by magnet experts have enabled a consensus to be built on choosing between some of the design options in the original ITER basic design in 2001. The major design decisions were to maintain the circular Nb3Sn conductor embedded in radial plates for the toroidal field (TF) coils and to maintain NbTi-based conductors for the PF coils. Cold testing, at low current, is also being introduced for quality control purposes for all coils. [Copyright &y& Elsevier]