Search

The Divertor Tokamak Test (DTT) vacuum vessel (VV) is a toroidal chamber that ensures an enclosed vacuum environment for the plasma, a confinement barrier, and temperature control. Water flows in the double-shell D-shaped VV cross-sectional structure made from stainless steel to operate at a set temperature. Beyond temperature control, the water cooling circuit needs to act as a neutron shielding media to protect the structures installed outside of the VV, namely, the superconducting coils. The shielding function is achieved due to the addition of boric acid in the water. The requirement on the borated water for the DTT VV is to have 8000-ppm B solution highly enriched in ¹⁰B (95% ¹⁰B). Given the lack of water chemistry guidelines for fusion power plants, the water chemistry requirements from fission power plants were investigated. In this work, general corrosion of stainless steel, 316L type family, in concentrated borated water solutions was investigated experimentally using metal release test. Samples were exposed to ultrapure water (UPW) and 8000-ppm B borated water at 80 °C for one week to quantify the amount of ions released in solution. 316L general corrosion was studied considering the different water chemistries, UPW versus borated water, and steels microstructures. DTT VV presents many welded joints, so general corrosion of welding-induced microstructure was here investigated compared to 316L base microstructure. The release of ions from general corrosion was found to be more influenced by water chemistry than microstructure.

This paper focuses on the development of the water-cooled divertor target concept known as Thermal Break, which was carried out in two phases. In Phase 1, six small scale mock-ups were fabricated and subjected to high heat flux (HHF) testing of up to 25 MW/m2 and thermal cycling of up to 500 cycles at 20 MW/m2. All six mock-ups survived the campaign and maintained 20 MW/m2 heat exhaust capability. Detailed examination of mock-ups was carried out to understand the damage mechanisms. One mock-up, which was tested beyond its design intent at 500 cycles, shows signs of progressive damage. Potential damage modes were identified and influenced subsequent Phase 2 mock-up design. Although there are signs of tungsten surface cracking, the predominant damage mode is not by “deep cracking” but substantial permanent deformation in the interlayer features. Therefore, in Phase 2 the manufacturing procedure was updated, the interlayer grooves were given stress-relieving radii which have significantly reduced the interlayer plastic strain range. Interlayer design parameters were selected following the use of response surface-based design search and optimization. Mock-ups of the Phase 2 design have been manufactured and HHF testing is planned within 2018.

The construction of a Divertor Tokamak Test (DTT) facility is fully included in the European effort towards the realization of fusion energy with the aim of studying possible alternative solutions, with respect to the ITER research program, to tackle the problem of power exhaust. This means that the machine must be compatible with different possibilities for the divertor cassette design and the plasma scenario. The design of first “reference” divertor is largely based on the ITER-like divertor, but in the future other solutions are going to be tested, including liquid metal divertors. To provide a justification basis for the design of the reference cassette and guidelines for the future experimentations it is fundamental to define the requirements of this power exhaust system, in particular the heat flux compatibility and the interfaces with the surrounding interacting systems of the tokamak (the water cooling pipes, the divertor magnetic coils, the divertor diagnostics, etc.). This work shows how the divertor system can be included in a wider Requirement Management procedure that has been defined for DTT highlighting the relevant interface parameters. Additionally, a data analysis tool has been developed to provide an interactive visualization of requirements traceability to directly show the impact of changes in the divertor on the design and operations of the machine and viceversa. In this way, during future studies on alternative design divertors or plasma scenarios, the limiting constraints on both sides can be easily individuated and traced to their motivation, and possibly discussed.

The Divertor Tokamak Test (DTT) facility is an experimental facility under design and construction at ENEA C.R. Frascati. The aim of DTT is to investigate the power exhaust problem in a tokamak, providing possible alternative divertor solutions, with respect the conventional one, which can be extrapolated to DEMO fusion reactor. One of the main components of DTT facility is the Vacuum Vessel (VV), which has the function of providing an enclosed, vacuum environment for the plasma, acting also as a first confinement barrier. Starting from geometrical constraints, imposed by the desired plasma scenario and the configuration needed for the magnetic coils, the conceptual design of the VV was developed and dedicated research activities were carried out to verify design choices. A multiphysical approach was adopted with the aim of identifying a feasible and reliable solution for the VV, taking into account functional and design requirements, relevant aspects and issues. In particular, to assess and to optimize the VV design, fluid-dynamic, thermal and structural analyses were carried out, which are presented in the paper, as well as the progresses of the design.

Power exhaust is a key mission in the roadmap to the realization of a future fusion reactor. Among the different solutions, the use of liquid metals as plasma facing materials are of interest due to their potential increased lifetime. Several liquid metal limiters have been successfully tested in the Frascati Tokamak Upgrade over the last 10 years. Liquid materials such as lithium and tin have been investigated using capillary porous systems (CPSs), and their impact on plasma performance has been explored. From such experience, a liquid metal divertor (LMD) concept design, CPS-based, is here proposed. Tin has been preferred as plasma facing material. The proposed LMD would operate, in low evaporative regime, with matching heat exhausting capabilities to those of the baseline ITER-like divertor. Continuous refilling of the CPS is guaranteed with a reservoir at the back of the unit, where the metal is kept liquid through a gas heating circuit. The study has been carried out using ANSYS and the thermal results will be shown. All the design choices are compatible with the current materials and the constraints adopted for the DEMO W divertor. Using such configuration, thermal loads up to 20 MW/m2 are exhausted while keeping the surface temperature below 1250 °C. The design foresees values of pressure, temperature and flow rate of the water coolant in the same range expected for the W DEMO divertor, thus facilitating the integration of such solution in the current cassette design. Technological and practical aspects are addressed, i.e. tin corrosion and CPS wettability. Possible solutions to prevent tin corrosion, and its compatibility with structural materials, will be outlined.

The development of a divertor target for DEMO is of great importance, being able to sustain the harsh environment that is imposed on this component. To fulfill the loading requirements, different concepts were developed within the EUROfusion WPDIV project. The baseline concept is based on the ITER divertor target W-monoblock design. It is made of tungsten as armour material, CuCrZr as structural material and Cu-OFHC as compliant layer. One of the proposed alternative concepts aims to minimize the stress at interfaces by replacing the thick copper interlayer by W/Cu functionally graded material (FGM). In this study, the FGM interlayer, with a thickness of 500 μm, is composed of stacked elementary layers with the following three compositions: 25 vol.%W + 75 vol.%Cu, 50 vol.%W + 50 vol.%Cu, 75 vol.%W + 25 vol.%Cu. Several FGM interlayers were studied. In total three monoblock type mock-ups were manufactured. This paper describes the steps needed to manufacture mock-ups and characterization of elementary layers (composition, porosity, Young’s modulus). HHF tests applying up to 1000 cycles at 20 MW/m² and sub-sequent post-mortem examinations were performed to qualify the concept performance.

The Divertor Tokamak Test facility is an Italian experimental facility under design and construction at ENEA C.R. Frascati. The main goal of DTT is to provide an integrated environment, relevant to DEMO, where testing possible solution to the power exhaust problem in a tokamak (like for example: i) Plasma facing components technology ; ii) Plasma and divertor shape; iii) impurity seeding to increase radiation). In this respect, DTT has been designed to be flexible and adopting technologies relevant to DEMO. After its initial inception in 2015, concluded with the publication of the "DTT project proposal", a complete re-baseline has been provided oncluded with the publication of the "DTT Interim Design Report" in 2019, aimed at accommodating the request of flexibility coming from the international fusion community. During 2019, the engineering integration activity has started and the first construction contracts have been signed. This paper provides an overview of the integrated design activity towards the realization of the facility within 2025.

The divertor is one of the most challenging components for DEMO reactor both from the design and fabrication technology point of view, since it must be capable to withstand the high heat fluxes (HHF) expected during normal operation (up to 10 MW/m2) and slow transient events (up to 20 MW/m2), like loss of plasma detachment. Within the frame of the EUROfusion Consortium the “Target development” subproject inside the Work Package ‘Divertor’ (WPDIV) has been dedicated to achieve this performance studying different concepts. ENEA focused on the “ITER-like” target that consists of applying the ITER design and fabrication technology to DEMO targets. The ITER-like concept mock-up was at first designed and optimized by FE analysis then manufactured and checked by non-destructive testing (NDT), finally thermal fatigue testing (TFT) and destructive testing (DT) were performed too. This paper reports the comparison of the thermal behavior of the ITER-like mock-ups between the analysis performed using ANSYS-CFX (Computational Fluid Dynamic analysis, CFD) and the outcome of the High Heat Flux tests (HHF) made at GLADIS facility. Finally the mechanical behavior of the mock-ups has been compared in terms of ratchetting applying dedicated criteria of the ITER SDC-IC and the fatigue lifetime has been estimated by means of Low Cycle Fatigue Curves (LCFC).

One of the most critical parts of a high heat flux plasma facing component of a Tokamak divertor is the armour to heat sink joint. R&D activity has been launched in the frame of the EUROFUSION Power Plant Physics & Technology programme and in particular in the Divertor project area in order to investigate the possibility of using the International Thermonuclear Experimental Reactor (ITER) design and fabrication technology in an EU-DEMO tokamak. ENEA is being involved in the R&D ITER activities since several years developing suitable joining technologies for divertor plasma facing components fabrication. In fact, Hot Radial Pressing technique was used for the manufacturing of several small and medium scale prototypes that have been successfully tested by e-beam thermal fatigue testing. A new furnace was designed and installed at ENEA labs with the objective to extend the Hot Radial Pressing (HRP) manufacturing technology to EU-DEMO relevant divertor plasma facing components. This new furnace uses the pipe heat sink of these components as heater, speeding up the heating phase and allowing a better control of the pipe temperature. Considering that in EU-DEMO the thermo-hydraulic operative conditions of these components are different from ITER a FEA parametric optimization was performed and the geometry of the small mock-up to be manufactured was defined accordingly. The six manufactured mock-ups were examined by Ultrasonic and SATIR non-destructive examination methods in order to check the quality of the joining before their delivering to the high heat flux tests facility GLADIS for the thermal fatigue testing at high heat flux.

Electrical insulation of superconducting magnets represents one of the most critical aspects in the manufacturing and operation of these devices. Accidents during manufacturing or severe out-of-design loadings in operations could compromise the design life of the component. In this respect, since they do not present a ductile behavior, when elastic limits are exceeded, insulating materials could be irreversibly damaged. Damages could be cracks or delaminations producing different effects in terms of operational life of the magnets. While a crack could give rise to electrical breakdown during fast discharge after a quench, delaminations might produce local heating due to friction that might induce superconducting quench. This paper shows the results obtained with two finite element models of a winding pack mock-up of a superconducting magnet for fusion applications subject to off-normal torsion conditions. The mockup is supposed to be made of a matrix of superconducting cable-in-conduit-conductors wrapped by 1 mm thick glass fiber resin composites. Through a detailed analysis of stress-strain properties of the insulation and a comparison with available experimental results, the authors give a threshold to the off-normal loading conditions under which the magnet could be safely operated.

We analyzed data from ultrasonic testing and infrared thermography non-destructive examinations performed on a subset of 25 small scale water-cooled target mockups of four target designs developed in the 2nd R&D phase of WPDIV. Examinations were performed before high heat flux tests for the 25 mockups and after high heat flux tests for 18 of them. The detected manufacturing defects are various in size and location. Widths of defects range from 0.2 to 12 mm, which is the entire block width. Angular sizes of defects range from few degrees to 360°. Defects having an angular size less than 50° or a width less than 4 mm are likely to be missed by infrared thermography examinations. After high heat flux tests at 20 MW m−2 up to 500 cycles, we noticed no significant evolution of pre-existing defects.

As a major in-vessel component of a tokamak-type fusion reactor, the divertor is mainly in charge of removal of particles and partial power exhaust via scrape-off layer. The target plate of the divertor is directly exposed to non-uniform high heat flux on the surface by particle bombardment and radiation. In the case of ITER and a DEMO reactor, the peak surface heat flux is expected to reach up to 10 MW/m2 during normal operation and 20 MW/m2 during slow transient events like loss of plasma detachment. This paper reports the results of a preliminary code-based design study and fabrication technology verification test which were conducted for developing an ITER-like divertor target design for the DEMO divertor. The structural failure evaluation against the ratchetting and fatigue criteria of the ITER SDC-IC showed that the design with reduced dimensions would allow sufficient design margin (reserve factor) for three distinct thermal loading cases. The first trial of mock-up fabrication using a new joining furnace at ENEA was successfully completed. The ultrasonic inspection test made before and after the cyclic HHF tests at GLADIS facility demonstrated high quality of fabrication and robust design concept.

ENEA and Ansaldo Nucleare S.p.A. are currently working on the fabrication of full scale prototypes of inner vertical target component for ITER (International Thermonuclear Experimental Reactor). The requirements is to manufacture a component by means of a qualified technology which uses armour and cooling pipe materials defined by ITER organization specification. A dedicated new Hot Radial Pressing facility was designed and finally installed in ENEA for the fabrication of the full scale Plasma Facing Units of the inner vertical target component prototype. HRP technology was widely tested with small mock-ups and medium qualification prototypes but its applicability to fabricate full scale units has to be anyway demonstrated and, therefore, their joint quality and also dimensions and shape has to be properly checked. The compliance of the joint quality with the acceptance criteria is then performed by ultrasonic water gap technique, using the qualified ENEA testing equipment. Dimensional check is performed as well by analyzing acquired data points compared to the reference drawings and 3D model.

One of the main objectives of the DTT project is to test many divertor designs and configurations, so that the concept of the machine could change from the initial single null (SN) configuration to other configurations such as the SnowFlake Divertor (SFD). Furthermore the design of Vacuum Vessel, ports and In-Vessel Components should take into account the application and testing of a Liquid Metal Divertor. For this reason the divertor design has been developed having in mind the possibility of easily replacing the divertor itself by remote handling.

In this paper, a comparison is presented among the ultrasonic measures of different W-monoblock mock-ups following high heat flux tests. The DEMO reference solution for the divertor targets (ITER-like) is compared with two other concepts, in which the interlayer between monoblocks and tube has been engineered to increase the performance of the component (Thermal Break Interlayer, Thin Graded Interlayer). Ultrasonic measurements have been performed at the depths of interest. The reflected signal amplitude and time of flight at defined depths are used to determine the shape and entity of the defect, when present. For each concept, the most frequently observed defects due to thermal loading are then identified. Preliminary post-mortem analysis confirms the results from the ultrasonic tests, proving the defect imaging capability of such non-destructive technique for advanced W-monoblock concepts.

International audience; For the development of a divertor for DEMO, the European WPDIV project is underway since 2014. The first phase of the project (2014–2016) aims to provide mock-ups adapted to DEMO operation requirements and the second phase aims to furnish mock-ups, with standardized geometry, which fulfill phase 1 requirements. Within the WPDIV project, several options are under development. One of these aims is to replace the thick copper interlayer, used for ITER divertor components, with a very thin coat (functional gradient material or pure copper) for armor-to-pipe joining. One of the benefits is related to armor temperature which is decreased as the distance of heat conduction path is shortened. Some blocks equipped with thin functional gradient material as interlayer proved, in 2016, to handle high cycling performances without any degradation (no surface change aspect and no decrease of thermal heat exhaust capability). This article gives a brief overview on the recent achievements of the development of thin interlayer concept focusing on the design, mock-up production, inspection, high heat flux (HHF) qualification testing and post-examinations.

ENEA has been working on improving the ultrasonic scanning technique of Plasma Facing Units (PFUs) components for many years, obtaining increasingly accurate and easy to interpret results. The first tests were performed on small mock-ups and, due to the geometry of the component, tests were performed from inside the tube. The first issue to solve was the reduced pipe thickness and the need to direct the ultrasound beam orthogonally to the tube surface. The choice of the correct probe characteristics and its particular shape solved this issue. Subsequently, the problem of a scan along a curved tube had to be addressed. Therefore, a special probe holder, able to follow the curvature placed at the end of a flexible shaft, was designed. The definition of more stringent acceptance criteria for the plasma facing components required to increase the scanning resolution. Defect detection and sizing were improved up to 0.5 mm by working on automatic system and data analysis software. The increasing of the sample sizes up to full-scale prototype for pre-production, led to the necessity to improve speed of the scanning without losing resolution. At this point, the scan along generatrix was discarded to pass to a helical path. In this work, the system created and patented in ENEA laboratories, for scanning full-scale Inner Vertical Target PFU prototypes of ITER, is illustrated. Test results in terms of resolution and defects sizing are shown.

The Divertor Tokamak Test (DTT) facility has been proposed by ENEA, in collaboration with other Italian institutions, to investigate power exhaust solutions with a machine capable to integrate all DEMO relevant physics and technology issues. The DTT machine will be able to host, in different phases of its life-time, advanced divertor magnetic configurations (snowflake, super-X, double null) and liquid metal solutions, and will be able to withstand the large loads expected in the DEMO fusion power plant. The first wall (FW) has been designed with coaxial pipes. It shall be compatible with liquid metal divertors and therefore be heated at a temperature above 200 °C to avoid metal condensation on the surface and, at the same time, withstand to thermal load up to 1.5 MW/m2 reached during the ramp-up. Two alternative conceptual designs are considered, one made of CuCrZr and one in stainless-steel, both coated with a W layer deposited by plasma spray (PS) technique. The possibility to use a FW without active cooling or cooled by gas or water was studied. In addition, a preliminary design optimization based on thermo-mechanical FEM analyses was carried out. To minimize the heat transferred to the superconductive coils, the vacuum vessel (VV), heated by water at 100 °C, is shielded by a Thermal Shield (TS) cooled with helium at 70 K. By means of a preliminary thermo-hydraulic analysis, the power requested to the cryogenic system has been estimated.

Fabrication and qualification of a representative full-scale prototype of the International Thermonuclear Experimental Reactor (ITER) divertor inner vertical target is the subject of the contract assigned to Ansaldo Nucleare S.p.A (ANN) by Fusion for Energy, the EU-Domestic Agency. ENEA, as major partner of this contract activities, was in charge of the plasma facing units (PFU) fabrication by means of Hot Radial Pressing. In parallel the other main fabrication steps were completed by ANN and its industrial partners based on the F4E technical requirements. Finally, the feasibility of the prototype assembly by integrating the PFUs onto the steel support structure has been demonstrated and the technical capability to reach the tight geometrical tolerances required for the plasma surface was checked. High heat flux thermal fatigue testing on the 8 full-W PFUs installed on the testing frame will take place at the ITER Divertor Test Facility at Efremov Institute in Saint-Petersburg to finally demonstrating the PFUs thermal removal capability. The reliability level reached by the HRP joining process and the industrial consolidation of the fabrication route, paves the way with confidence to a forthcoming series production of the ITER full-W divertor. The paper reports the full scale prototype main production achievements.

The divertor target plates are the most thermally loaded in-vessel components in a fusion reactor where high heat fluxes are produced on the plasma-facing components (PFCs) by intense plasma bombardment, radiation and nuclear heating. For reliable exhaust of huge thermal power, robust and durable divertor target PFCs with a sufficiently large heat removal capability and lifetime has to be developed. Since 2014 in the framework of the preconceptual design activities of the EUROfusion DEMO project, integrated R&D efforts have been made in the subproject ‘Target development’ of the work package ‘Divertor’ to develop divertor target PFCs for DEMO. Recently, the first R&D phase was concluded where six (partly novel) target PFC concepts were developed and evaluated by means of non-destructive inspections and high-heat-flux fatigue testing. In this paper, the major achievements of the first phase activities in this subproject are presented focusing on the design rationales of the target PFC concepts, technology options employed for small-scale mock-up fabrication and the results of the first round high-heat-flux qualification test campaign. It is reported that the mock-ups of three PFC concepts survived up to 500 loading cycles at 20 MW/m² (with hot water cooling at 130 °C) without any discernable indication of degradation in performance or structural integrity. Keywords: DEMO, Divertor targets, High-heat-flux, Plasma-facing component, Composites, EUROfusion

Fusion advanced study torus (FAST) has been designed with the aim to tackle the power exhaust problem with ITER and DEMO relevant bulk plasma. Considerable analysis effort has been spent in EM (electromagnetic) designing of FAST divertor components capable of withstanding the electromagnetic loads expected from the foreseen FAST operative conditions. Plasma disruption EM loads are one of the most critical load conditions for the divertor. Consequently a first dimensioning of the divertor for EM loads is mandatory. The foreseen FAST divertor structure is compact and the configuration is aimed to give to the structure the required mechanical and thermal capability as well as being remote handling compatible. The conceptual EM design of the divertor has been designed for FAST operative conditions and for DEMO relevant "Snow Flake" (SF) magnetic topology. Recently a "Snow Flake" (SF) magnetic topology has been suggested for the divertor region, capable to spread the plasma power flow on a much wider areas, with the possibility of reducing by a factor 4 the power flowing to the divertor tiles [1]. The conceptual design of the divertor has been obtained during activities based on the estimation of EM loads due to a typical Plasma Fast Down disruption event in FAST for the normal configuration and for SF magnetic topology. © 2013 Elsevier B.V.

The Fusion Advanced Study Torus (FAST), with its compact Tokamak design, high toroidal field and plasma current, will face many of the problems that ITER will meet and will anticipate many DEMO relevant physics and technology issues. The Design Upgrade of the Vessel and In-Vessel Components is presented in this paper. Relevant modifications were performed to the Vacuum Vessel (VV) and to the Plasma Facing Components (PFCs), i.e. the First Wall (FW) and the Divertor. The VV was modified to insert active reduction coils (ARC), between VV and the toroidal field (TF) coils to keep toroidal field magnet ripple lower than 0.3% and to allow Remote Handling for the FW and the Divertor. The FW, was modified to house coils for ELMs control and other plasma instabilities. A 3D thermo-hydraulic analysis using ANSYS code was performed to check FW heat removal capability. A new Divertor was designed to withstand the largest thermal loads of the high performance, low density, H-mode and to be able to comply with a recent magnetic topology called as "Snow Flake", increasing up a factor 4 the flux expansion. An exhaustive 3D thermo-hydraulic analysis using ANSYS code was carried out to show the capability of the Divertor to comply these high requirements. Design criteria were satisfied by present components of the upgraded machine. © 2013 Elsevier B.V.

The paper discusses a potential approach of infrared (IR)-analysis for high heat flux quality assessment of divertor tungsten monoblock components. The aim of these activities is the development of a statistics based method for the general application to the high heat flux testing of industrially manufactured W monoblock plasma-facing components. We have performed tests with 100 cycles at 10 MW/m 2 on 13 components or mock-ups manufactured for ITER, for the DEMO divertor development and for the WEST project. For eight W monoblock components with grinded surfaces, the statistical high heat flux (HHF) quality assessment method originally developed for the series manufacturing of carbon fibre reinforced carbon flat tile divertor components for Wendelstein 7-X was applied. The large variability of the emissivity of W as well as the monoblock geometry require a careful emissivity correction of IR camera data. The emissivity is strongly dependent on the W surface quality and final machining and influences the apparent temperature measured by an IR camera. Therefore we have applied a pragmatic correction algorithm based on two-colour pyrometry. Finally we propose a possible HHF testing strategy of a large number of industrially manufactured W monoblock components.

The future DEMO divertor is currently under conceptual design within the European Consortium. In this regard, several concepts have been proposed and mock-ups have been fabricated to investigate their thermo-mechanical behaviour. Indeed, as a key plasma facing component, the divertor will have to withstand extreme thermal loads (up to 20 MW m-2 during slow transient events) and will have to be able to exhaust a large amount of heat. The presence of structural defects in the component may significantly affect the thermal response and must therefore be considered. A non-destructive technique based on infrared thermography is proposed here to detect defects in mock-ups where graded material was used as an interlayer between the heatsink material and the armor material. Two methods to characterize the size and location of such defects are presented. It was shown that finite element analysis combined with experimental data from infrared thermography, provides accurate means to assess quantitatively the size and position of thermal imperfections. © 2017 Franklin Gallay.

ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m2 with the cooling water of 2 m/s, 20 C, 0.2 MPa. © 2013 Elsevier B.V.

ENEA and Ansaldo Nucleare S.p.A. have been deeply involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities for the manufacturing of high heat flux plasma-facing components (HHFC), and in particular for the inner vertical target (IVT) of the ITER divertor. This component has to be manufactured by using both armour and structural materials whose properties are defined by ITER. Their physical properties prevent the use of standard joining techniques. The reference armour materials are tungsten and carbon/carbon fibre composite (CFC). The cooling pipe is made of copper alloy (CuCrZr-IG). During the last years ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components of different length, geometry and materials, by using innovative processes: HRP (hot radial pressing) and PBC (pre-brazed casting). The history of the technical issues solved during the R&D phase and the improvements implemented to the assembling tools and equipments are reviewed in the paper together with the testing results. The optimization of the processes started from the successful manufacturing of both W and CFC armoured small scale mockups thermal fatigue tested in the worst ITER operating condition (20 MW/m2) through the achievement of record performances obtained from a monoblock medium scale mockup. On the base of these results ENEA-ANSALDO participated to the European programme for the qualification of the manufacturing technology to be used for the procurement of the ITER divertor IVT, according to the F4E specifications. A divertor inner vertical target prototype (400 mm total length) with three plasma facing component units, was successfully tested at ITER relevant thermal heat fluxes. Now, ANSALDO and ENEA are ready to face the challenge of the ITER inner vertical target production, transferring to an industrial production line the experience gained in the development, optimization and qualification of the PBC and HRP processes. © 2013 Elsevier B.V. All rights reserved.

ENEA and Ansaldo Nucleare S.p.A. (EA) have been deeply involved in the European International Thermonuclear Experimental Reactor (ITER) research and development activities for the manufacturing of ...

FAST (Fusion Advanced Studies Torus) is a proposal for a Satellite Facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploit innovative plasma facing component systems for DEMO. FAST is a compact (Ro = 1.82 m, a = 0.64 m, triangularity δ = 0.4) and cost effective machine able to investigate, with integration capability, non linear dynamics effects of alpha particle behaviour in burning plasmas. FAST operates in high performance H-mode (BT up to 8.5 T; IP up to 8 MA), as well as in advanced tokamak mode (IP = 3 MA), and in full non inductive current mode (IP = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets. This allows for a pulse duration up to 170 s at 3 MA/3.5 T. The vacuum vessel (VV), segmented into 20-degree modules, is capable to accommodate a 40 MW RF power system. The machine has been designed to house a 10 MW Negative Neutral Beam Injection (NNBI) system. Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plate materials, and argon or neon as the impurities to be injected for mitigating the thermal loads.

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets. In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes. According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m2 and finally 1000 cycles at 20 MW/m2. After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing. A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area. The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks. This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.

This paper discusses the application of non-destructive testing (NDT) by ultrasonic technique for the control of the joining interfaces of the ITER divertor vertical target plasma facing units. The defect detection capability has to be proved for both metal to metal and metal to carbon/carbon fibre composite (CFC) joints because these two types of joints have to be realized for the manufacturing of the high heat flux units. In this paper the UT results coming from the investigation performed during the manufacturing, but also after the thermal fatigue testing (up to 20 MW/m 2 ) of six mock-ups manufactured using the Hot Radial Pressure technology (HRP) in ENEA labs are presented and compared with the evidences from the final destructive examination. Regarding the Cu/CFC joint, the effectiveness of the ultrasonic test has been deeply studied due to the high acoustic attenuation of CFC to ultrasonic waves. To investigate the possibility to use the ultrasonic technique for this type of joint, an ‘ad hoc’ flat Cu/CFC joint sample, that reproduces the actual annular joint interfaces, was manufactured. This flat sample has the advantage of being easily tested by probes with different geometry and frequency. UT results are compared with X-ray and eddy current testing of the same sample.

The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint. An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs. ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies. According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study. After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects. In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor. This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results.

In the present study a complete neutronic analysis has been performed for the current design of the JT-60SA toroidal field coil (TFC) system. The MCNP5 Monte Carlo code has been used to calculate the nuclear heating, neutron spectra and absorbed dose in the TFC components, assuming a DD neutron emission rate of 1.5 × 1017 n/s (and 1% DT). Nuclear heating of the winding pack is lower than 0.3 mW/cm3 and the maximum nuclear heating of the TFC case is 0.4 mW/cm3. The overall nuclear heating, including the safety margin, is less than 8 kW. Spatial distribution of the nuclear heating has been provided along poloidal, radial and toroidal directions as to be used for thermo-hydraulic analysis and the design of TFC system. The absorbed dose to insulator is as low as to avoid the replacement during the whole life of the machine. Neutron fluxes have been used as input for a preliminary activation analysis performed with FISPACT inventory code. Activity and contact dose rates have been calculated at different cooling times, after 10 years of operations in some representative zone of the winding pack and the case. All the TFC materials can be easily recycled within the first day after shutdown and the hands-on recycling is possible within less than 30 years.

Manufacturing of prototypical components by means of the two ENEA patented technologies, namely PBC (Pre-Brazed Casting) and HRP (Hot Radial Pressing) was successfully concluded in the frame of EFDA contracts. The successful results of the thermal fatigue testing performed at FE200 (200 kW electron beam facility, CEA/AREVA France) on these components are beyond ITER requirements (3000 cycles at 10 MW/m2 for 10 s on both CFC and W part, then 2000 cycles at 20/15 MW/m2 for 10 s on CFC/W part, respectively) and they have confirmed that the developed technologies are candidates for the manufacturing of monoblock divertor components. In order to reach these successful results a reliable non-destructive testing (NDT) procedure was set up. The chosen method was the ultrasonic water gap technique. This technique was chosen mainly for its reliability and because it can be applied during each step of the manufacturing route. The capability to detect any defect zones and to determine its actual position and extension was achieved by designing and assembling a suitable ultrasonic equipment. A specific software, developed at ENEA, is able to give all the required information. The NDT procedure was qualified by manufacturing several small samples containing reference defects of given size and position and by verifying the system capability to detect them. The destructive examination performed on some samples and on divertor vertical target prototypes has confirmed the reliability of the NDT equipment and procedure.

The upgrade of JT-60U to JT-60 Super Advanced (JT-60SA), a fully superconducting tokamak, will be performed in the framework of the Broader Approach (BA) agreement between Europe (EU) and Japan. In particular, the Toroidal Field (TF) system, which includes 18 coils, is foreseen to be procured by France, Italy and Germany. This work covers activities from design and manufacturing to shipping to Japan. The present paper is mainly devoted to the analyses that lead to the conductor design and to the technical specifications of the joints for the JT-60SA TF coils. The conductor geometry is described, which is derived from Cable-In-Conduit concept and adapted to the actual JT-60SA tokamak operating conditions, principally the ITER-like scenario. The reported simulations and calculations are particularly dealing with the stability analysis and the power deposition during normal and off-normal conditions (AC losses, nuclear heating). The final conductor solution was selected through a trade-off between scientific approach and industrial technical orientation. Besides, the TF system connections layout is shown, derived from the industrially assessed twin-box concept, together with the associated thermo-hydraulic calculations ensuring a proper temperature margin.

In the framework of the IFMIF R&D program, a design-integration and a thermo-mechanical analysis of the Li target backplate has been carried out. The stress and displacement fields, due to the heating induced by the neutron flux, have been computed with the aim to check if the stress remains below the material's allowable values and the induced displacement field leads to an acceptable deformation of the surface where the lithium flows, so that no flux instabilities are induced in the lithium jet. The analysis has been performed for two different backplate geometries, namely a thick and a thin option, for two different thermal and mechanical constraint cases and for different materials: austenitic steel 316L, IEA F82H steel and Mod 9Cr1Mo steel (T91). The results, obtained by means of ABAQUS code, show that AISI 316L steel might be not suitable as black-plate material, conversely, both the F82H and the T91 steels satisfy ASME requirements. The thin plate solution appears preferable in order to reduce the stress field.

The objective of the present work was to characterize commercial-grade Tyranno SA SiC fiber reinforced chemically vapour infiltrated (CVI) SiC matrix composites (SiCf/SiC) with chemically vapour deposited (CVD) SiC coating. The characterization includes the assessment of the monotonic mechanical properties. Low cycle flexural fatigue behaviour has been investigated at room temperature (RT) and 1000 °C by means of 4-point bending tests. The creep behaviour at 1000 °C was preliminary investigated by means of constant bending stress rupture test. The material showed a pronounced degradation of monotonic mechanical properties at high temperature. Low cycle flexural fatigue behaviour showed excellent and satisfactory results at RT and 1000 °C, respectively. The creep resistance at 1000 °C is significant only at low load level.

ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European International Thermonuclear Experimental Reactor (ITER) development activities for the manufacturing of the inner vertical target (IVT) plasma-facing components of the ITER divertor. © 2015 IEEE.

Divertor is a crucial component in Tokamaks, aiming to exhaust the heat power and particles fluxes coming from the plasma during discharges. This paper focuses on the optimization process of FAST divertor, aimed at achieving required thermo-mechanical capabilities and the remote handling (RH) compatibility. Divertor RH system final layout has been chosen between different concept solutions proposed and analyzed within the principles of Theory of Inventive Problem Solving (TRIZ). The design was aided by kinematic simulations performed using Digital Mock-Up capabilities of Catia software. Considerable electromagnetic (EM) analysis efforts and top-down CAD approach enabled the design of a final and consistent concept, starting from a very first dimensioning for EM loads. In the final version here presented, the divertor cassette supports a set of tungsten (W) actively cooled tiles which compose the inner and outer vertical targets, facing the plasma and exhausting the main part of heat flux. W-tiles are assembled together considering a minimum gap tolerance (0.1-0.5 mm) to be mandatorily respected. Cooling channels have been re-dimensioned to optimize the geometry and the layout of coolant volume inside the cassette has been modified as well to enhance the general efficiency. © 2015 Elsevier B.V. All rights reserved.

ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5-10 MW/m2 but the capability to remove up to 20 MW/m2 during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. The series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative 'dummies' in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU. © 2015 Elsevier B.V. All rights reserved.

A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under 'DEMO' conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an "optimized" ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m-2, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes. © 2015 Elsevier B.V. All rights reserved.

In order to prevent the overheating of the liquid Li surface and the consequent Li evaporation for T > 500°C, an advanced version of the liquid lithium limiter has been realized and installed on FTU. This new system, named Cooled Lithium Limiter (CLL), has been optimized to demonstrate the lithium limiter capability to sustain thermal loads as high as 10 MW/m2 with up to 5 s of plasma pulse duration. The CLL operates with an actively cooled system with water circulation at the temperature of about 200°C, for heating lithium up to the melting point and for the heat removal during the plasma discharges. To characterize CLL during discharges, a fast infrared camera and the spectroscopic signals from Li and D atom emission have been used. The experiments analyzed so far and simulated by ANSYS code, point out that heat loads as high as 2 MW/m2 for 1.5 s have been withstood without problems. © 2014 Elsevier B.V.