Contents lists available at ScienceDirect Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes Analysis of stress induced by plasma disruption on vacuum vessel through multi-physics modeling V. Cocilovo a, *, R. Fresa b a ENEA-FSN-FUSTEC, c.r.e. Frascati, via E. Fermi, 45 00044, Frascati (ROMA), Italy b Scuola di Ingegneria- Università della Basilicata, Via dell’Ateneo Lucano, 10. 85100, Potenza, Italy ARTICLE INFO Keywords: Stress analysis Vacuum vessel VDE Disruption Tokamak ABSTRACT The analysis of the stresses induced on the vacuum vessel (VV) of a Tokamak and its internal components by the plasma instabilities, such as plasma disruptions, also following a Vertical Displacement Event (VDE), is one of the major concern in the Tokamak mechanical design. So the availability of a fast simulation tool for evaluating the different design options and also able to perform parametric analysis is highly attractive to define the project requirements and to verify the conceptual design. To respond to this wish, a methodology based on the multi- physics modeling capability offered by the Comsol® software platform was developed. It consists in coupled simulations based on the data sharing between two 2D axisymmetric models, everyone coupled with two cor- responding 3D models shifted each other of 10 degrees. In the 2D axisymmetric model of every couple, the magnetic and electric fields generated by the plasma VDE and disruption are calculated imposing the plasma time evolution as input. In the corresponding 3D model only the metallic structures are present and on them the Electric and Magnetic fields are extruded, determining so the induced currents diffusion in the passive conductors. Then the resulting Lorentz’s forces are imposed as body loads on the mechanical structures, so a linear stress analysis can be carried out after the constraints assignment. The same procedure is followed with the second couple of 2D/3D models for check purposes, comparing some proper physical quantities, such as the total induced current flowing on the VV. In this paper the methodology is presented by reporting the simulation of a double null plasma VDE, lasting c.a. 100 ms, followed by a full 5.5 MA plasma current quench in about 40 ms, in a medium size Tokamak intended for experimental research purposes. 1. Introduction Plasma disruptions, following or not a Vertical Displacement Event (VDE), are the most dangerous event for the mechanical structures of a Tokamak machine, especially if it happens at high plasma current. Their effects can generate mechanical loads such to damage seriously the internal components in the Vacuum Vessel (VV) and the VV itself. So they are a big concern for the Tokamak designer since the early steps of a new project. Great efforts have been done in the fusion community to deal with the plasma – mechanical structures interaction and a huge amount of methods have been developed (see ref. [1] for an extensive biblio- graphy). But the main aim of many of these works is direct to the plasma evolution analysis for stability check and control of the plasma parameters. The valuable results obtained with these methods are however confined in the electro-magnetic fields and then they must be transferred in ad hoc mechanical codes to evaluate if the design under check is compatible with the mechanical stresses arising from the Lorentz’s forces on the metallic structures of a Tokamak. This data transfer between different software codes can be a cumbersome, un- standardized and not error free process. Differently our approach aims to get a fast, integrated and agile simulation tool to verify and optimize conceptual design of mechanical structures surrounding the plasma (VV and in vessel components) since the early stage of the design process, even to compare different design solutions. This point of view, based on engineering instances, doesn’t need an accurate simulation of plasma-surrounding conductive struc- tures interaction, but instead a well defined geometry of the metallic structure to evaluate on them the induced currents path and density, and therefore the consequent body loads so to make possible a reliable stress analysis. The plasma evolution, displacement and current quench, evaluated by dedicated codes such as the one presented in ref. [1] or derived by experimental data collected on existing Tokamaks, is imposed as input to the model. In this way are disregarded in the https://doi.org/10.1016/j.fusengdes.2020.111684 Received 13 September 2019; Received in revised form 4 April 2020; Accepted 5 April 2020 ⁎ Corresponding author. E-mail address: valter.cocilovo@enea.it (V. Cocilovo). Fusion Engineering and Design 157 (2020) 111684 0920-3796/ © 2020 Elsevier B.V. All rights reserved. T