102,495 research outputs found
Electromagnetic disruption analysis in IGNITOR
No full textThe present paper reports a detailed analysis of disruptions in the IGNITOR tokamak. Two different numerical models are used (one axisymmetric, the other three-dimensional), in order to validate the results and to highlight specific effects due to different assumptions. The results show a good agreement between the two codes, when similar assumptions are made, and highlight the potential significance of three-dimensional effects. © 2015 Elsevier B.V
Modeling of Electromagnetic Loads on DEMO First Stage Divertor
No full textIn this work the analysis of the effects of the poloidal currents flowing on the cooling piping of the divertor armour tiles is carried out. To deal with the complexity of the problem a parametric solving scheme, starting from the nominal plasma current value, was adopted to contemplate the great variability of the possible cases deriving from the experimental data base and to compensate the lack of knowledge due to the not well assessed theory on the plasma wall interaction. Further to overcame the difficulties in modeling the real design of the piping with the necessary spatial resolution to individuate the local current concentration areas the methodology illustrated here is based on shell interfaces for solving either the electric and the mechanical problem. This approach proved to be capable to highlight the critical design areas and was useful to suggest the relative remedial corrections. © American Nuclear Society
Modelling of the Ignitor scrape-off layer including neutrals
No full textIgnitor is a tokamak project aimed at achieving ignition. In the reference scenario, plasma-surface interactions are controlled by a Mo first-wall/limiter, which constitutes a simple engineering solution but, at the same time, a special challenge for edge plasma modelling. Here the ASPOEL plasma fluid code, already applied to Ignitor in the recent past, is coupled with the neutral Monte Carlo code EIRENE. We study the effects of the neutrals on the plasma density and temperature profiles in the Ignitor scrape-off layer, and compute the particle and heat loads onto the Ignitor first-wall limiter
Modeling of long term effects on plasma facing components for a fusion power reactor
No full textIn a fusion power reactor the Plasma Facing Components (PFC) will experience a thermal and neutron irradiation induced creep together with tensile properties degradation and swelling due to neutron irradiation. So the investigation of the long term creep effects on the materials used for the PFC's in a fusion power plant are of vital importance for the design and safe operation of the device. On the other hand the creep behavior study for a given material requires long and expensive test campaigns, repeated on specimens at different levels of neutron irradiation, because of the material parameters variation due to the cumulated irradiation. In this work we want to investigate if the numerical mechanical simulations employment, according to a proper methodology, could reduce the number of needed creep tests, because this would be a valuable help in defining suitable materials and valid conceptual designs for PFC's. For this reason a method based on the systematic variation of the parameters of the empirical law, e.g. the Norton-Bailey, is outlined. To exemplify it, the behavior of a simplified model is analyzed under thermal and mechanical cyclic loading in a time transient elasto-plastic simulation, including the creep behavior, varying the parameters in the empirical creep law for the material
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
Experiments and modeling on FTU tokamak for EC assisted plasma start-up studies in ITER-like configuration
No full textThe intrinsic limited toroidal electric field (0.3 V m-1) in devices with superconducting poloidal coils (ITER, JT-60SA) requires additional heating, like electron cyclotron (EC) waves, to initiate plasma and to sustain it during the burn-through phase. The FTU tokamak has contributed to studying the perspective of EC assisted plasma breakdown. Afterward, a new experimental and modeling activity addressing the study of assisted plasma start-up in a configuration close to the ITER one (magnetic field, oblique injection, and polarization) has been performed and is presented here. These experiments have been supported by a 0D code, BKD0, developed to model the plasma start-up and linked to a beam tracing code computing, in a consistent way, EC absorption. The FTU results demonstrate the role of polarization conversion at the inner wall reflection. Dedicated experiments also showed the capability of EC power to sustain plasma start-up in the presence of strong error field (12 mT), with a null outside the vacuum vessel. The BKD0 code, applied to FTU data, has been used to determine the operational window of sustained breakdown as a function of toroidal electric field and neutral pressure. Experimental results in agreement with the BKD0 simulations support the use of the code to predict start-up in future tokamaks, like ITER and JT60SA. © 2015 EURATOM
First disruption studies and simulations in view of the development of the DEMO Physics Basis
No full textIn the development of the DEMO Physics Basis an important role is played by the prediction of the plasma disruption features and by the evaluation of the electro-magnetic (EM) and thermal loads associated with these events. Indeed, the kind and number of foreseen plasma disruptions drive the development of the DEMO operation scenarios and the design of vessel and in-vessel components. To characterize a plausible macroscopic plasma dynamics during these events, we will carry out an extrapolation from present-day machines of the main parameters characterizing the disruptions: thermal and current quench time, evolution of plasma current, β and li, safety factor limits, halo current fraction and width, radiated heat fraction. In particular, we will focus on extrapolations for the thermal and current quench characteristic times, due to their importance for the subsequent simulations aimed at the evaluation of the EM and thermal loads. The different options for DEMO design will be taken into account and the possible range of variation of the parameters will be estimated. The 2D axysimmetric MAXFEA and the 3D CarMa0NL codes will be used to evaluate the effects of the induced currents and the EM loads during a disruptive event and to analyze the various design options obtained by the PROCESS code. The results of these simulations, modeled as worst expected events, will be used as input for the system level analysis and design of the vessel and relevant in-vessel components. First simulations with CarMa0NL code including 3D structures are here presented to show the significant effects due to the large access ports. © 2015 Elsevier B.V. All rights reserved
Using MAXFEA code in combination with ANSYS APDL for the simulation of plasma disruption events on EU DEMO
No full textPlasma disruptions are one of the major concerns in the design phase of fusion devices. The very high eddy and Halo currents, induced in the passive structures, crossing the electromagnetic field generate huge loads. A Vertical Displacement Event (VDE) begins with a loss of position control that could be triggered by a plasma perturbation (e.g. ELMs, L-H and H-L transitions, minor disruptions, etc.) acting as a source of vertical and horizontal plasma displacements. At a certain point a fast Thermal Quench (TQ) takes place. After that the plasma current abruptly decreases (Current Quench (CQ) phase). During the plasma evolution, especially during the TQ and CQ, toroidal and poloidal eddy currents are induced in the metallic components, respectively due to the dynamic effect of plasma Poloidal Field Variation (PFV) and Toroidal Field Variation (TFV). The plasma time evolution and the effects of such events on the passive structures can estimated through 2D axisymmetric codes, such as MAXFEA. However, the presence of 3D structures (e.g. ports, divertor, etc.) generates non-trivial currents paths and distribution of EM loads. In order to estimate the 3D effects, MAXFEA has been used in combination with ANSYS APDL code, allowing to estimate both PFV and TFV consequences on the 3D model. Considering the DEMO PMI configuration and a fast upper Vertical Displacement Event (VDE), the procedure was successfully benchmarked, comparing the MAXFEA and APDL results, in a case where the 3D Vacuum Vessel (VV) was considered axisymmetric. The methodology has been then exploited and applied to estimate the EM load distribution on the real DEMO VV
EAST alternative magnetic configurations: Modelling and first experiments
No full textHeat and particle loads on the plasma facing components are among the most challenging issues to be solved for a reactor design. Alternative magnetic configurations may enable tokamak operation with a lower peak heat load than a standard single null (SN) divertor. This papers reports on the creation and control of one of such alternatives: a two-null nearby divertor configuration. An important element of this study is that this two-null divertor was produced on a large superconducting tokamak as an experimental advanced superconducting tokamak. A preliminary experiment with the second null forming a configuration with significant distance between the two nulls and a contracting geometry near the target plates was performed in 2014. These configurations have been designed using the FIXFREE code and optimized with CREATE-NL tools and are discussed in the paper. Predictive edge simulations using the TECXY code are also presented by comparing the advanced divertor and SN configuration. Finally, the experimental results of ohmic and low confinement (L-mode) two-null divertor and SN discharges and interpretative two-dimensional edge simulations are discussed. Future experiments will be devoted to varying the distance between the two nulls in high confinement (H-mode) discharges. © 2015 EURATOM
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