1,721,161 research outputs found
Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies
No full textITER-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
Quantitative thermal imperfection definition using non-destructive infrared thermography on an advanced DEMO divertor concept
No full textThe 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
Engineering of a FGM Interlayer to Reduce the Thermal Stresses Inside the PFCs
Full text linkFeatured Application: Thermomechanical design of a plasma-facing component with an engineered interlayer made with a functionally graded material, able to reduce the thermal stresses due to the mismatch of the coefficient of thermal expansion between the armor and the heat sink material. A substantial contribution of the stresses that arise inside the Plasma-Facing Components (PFCs) when a heat load is applied is caused by the mismatch of the Coefficient of Thermal Expansion (CTE) between the armor, usually made of tungsten (W), and the heat sink. A potential way to reduce such contribution to the secondary stresses is the use of an interlayer made with a Functionally Graded Material (FGM), to be interposed between the two sub-components. By tailoring the W concentration in the volume of the FGM, one can engineer the CTE in such a way that the thermal stresses are reduced inside the PFC. To minimize and, theoretically, reduce to zero the stresses due to the CTE mismatch, the FGM should ensure kinematic continuity between the armor and the heat sink, in a configuration where they deform into exactly the shape they would assume if they were detached from each other. We will show how this condition occurs when the mean thermal strain of each sub-component is the same. This work provides a methodology to determine the thickness and the spatial concentration function of the FGM able to ensure the necessary kinematic continuity between the two sub-components subjected to a generic temperature field monotonously varying in the thickness, while remaining stress-free itself. A method for the stratification of such ideal FGM is also presented. Additionally, it will be shown that the bending of the PFC, if allowed by the kinematic boundary conditions, does not permit, at least generally, the coupling of the expansion of the armor and of the heat sink. As an example of our methodology, a study case of the thermomechanical design of a steel-based PFC with an engineered W/steel FGM interlayer is presented. In such an exercise, we show that our procedure of engineering a FGM interlayer is able to reduce the linearized secondary stress of more than 24% in the most critical section of the heat sink, satisfying all the design criteria
Innovative design for FAST divertor compatible with remote handling, electromagnetic and mechanical analyses
No full textDivertor 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
European divertor target concepts for DEMO: Design rationales and high heat flux performance
No full textThe 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. © 201
Technological review of the HRP manufacturing process R&D activity
No full textENEA 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
Design study of ITER-like divertor target for DEMO
No full textA 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
Preliminary electromagnetic, thermal and mechanical design for first wall and vacuum vessel of FAST
No full textThe fusion advanced study torus (FAST), with its compact design, high toroidal field and plasma current, faces many of the problems met by ITER, and at the same time anticipates much of the DEMO relevant physics and technology. The conceptual design of the first wall (FW) and the vacuum vessel (VV) has been defined on the basis of FAST operative conditions and of "Snow Flakes" (SF) magnetic topology, which is also relevant for DEMO. The EM loads are one of the most critical load components for the FW and the VV during plasma disruptions and a first dimensioning of these components for such loads is mandatory. During this first phase of R&D activities the conceptual design of the FW and VV have been assessed estimating, by means of FE simulations, the EM loads due to a typical vertical disruption event (VDE) in FAST. EM loads were then transferred on a FE mechanical model of the FAST structures and the mechanical response of the FW and VV design for the analyzed VDE event was assessed. The results indicate that design criteria are not fully satisfied by the current drawing of the VV and FW components. The most critical regions have been individuated and the effect of some geometrical and material changes has been checked in order to improve the structure. © 2015 Elsevier B.V. All rights reserved
Further finite element structural analysis of FAST Load Assembly
No full textThe FAST (Fusion Advanced Study Torus) machine is a compact high magnetic field tokamak, that will allow to study in an integrated way the main operational issues relating to plasma-wall interaction, plasma operation and burning plasma physics in conditions relevant for ITER and DEMO. The present work deals with the structural analysis of the machine Load Assembly for a proposed new plasma scenario (10 MA - 8.5 T), aimed to increase the operational limits of the machine. A previous paper has dealt with an integrated set of finite element models (regarding a former reference scenario: 6.5 MA - 7.5 T) of the load assembly, including the Toroidal and Poloidal Field Coils and the supporting structure. This set of models has regarded the evaluation of magnetic field values, the evaluation of the electromagnetic forces and the temperatures in all the current-carrying conductors: these analysis have been a preparatory step for the evaluation of the stresses of the main structural components. The previous models have been analyzed further on and improved in some details, including the computation of the out-of-plane electromagnetic forces coming from the interaction between the poloidal magnetic field and the current flowing in the toroidal magnets. After this updating, the structural analysis has been executed, where all forces and temperatures, coming from the formerly mentioned most demanding scenario (10 MA - 8.5 T) and acting on the tokamak's main components, have been considered. The two sets of analysis regarding the reference scenario and the extreme one have been executed and a useful comparison has been carried on. The analyses were carried out by using the FEM code Ansys rel. 13. © 2013 Euratom-ENEA Association sulla Fusione
Acceptance tests of ITER vertical target divertor full scale plasma facing units fabricated by HRP
No full textENEA 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
- …
