Plasma Disruption Events Research Articles

Comparison between 3D eddy current patterns in Tokamak in-vessel components generated by disruptions

During plasma disruption events in Tokamaks, a large amount of magnetic energy is associated to the transfer of plasma current into eddy currents in the passive structures. In the ITER programme two design concepts have been proposed. One approach (ITER CDA design) is based on copper stabilization loops (i.e. twin loops) attached to box-shaped blanket segments, electrically and mechanically separated along the toroidal direction. For another design concept (ITER EDA design) based on lower plasma elongation there is no need for specific stabilization loops. The passive stabilization is obtained by toroidally continuous components (i.e. the plasma facing wall of the blanket segments allows a continuity along the toroidal direction). Consequently, toroidal currents flow, when electromagnetic transients occur. Electromagnetic loads appear in the blanket structures in case of plasma disruptions and/or vertical displacement events either for the ITER CDA design concept or for the ITER EDA design concept. In this paper the influence of the in-vessel design configuration concepts-insulated segments or electrically continuous structures-in terms of magnetic shielding and electric insulation on the magnitude and the flow pattern of the eddy currents is investigated. This investigation will allow a performance evaluation of the two proposed design concepts.

ELECTROMAGNETIC EFFECTS ON TOKAMAK IN‐VESSEL COMPONENTS IN PLASMA OFF NORMAL CONDITIONS

During plasma disruption events in tokamaks, a large amount of magnetic energy is associated to the transfer of plasma current in eddy currents into the passive structures. In the paper, the influence of the in‐vessel design, in terms of magnetic shielding and electrical insulation, on the magnitude and the flow pattern of the eddy currents is investigated.

Impact of the passive stabilization system on the dynamic loads of the ITER first wall/blanket during a plasma disruption event

Impact of the passive stabilization system on the dynamic loads of the ITER first wall/blanket during a plasma disruption event

Impact of the passive stabilization system on the dynamic loads of the ITER first wall/blanket during a plasma disruption event

Abstract In next-generation tokamak devices (i.e. ITER), passive stabilization of the plasma is required to mitigate the consequences of the plasma vertical displacements and to reduce the occurrence of plasma disruptions. With this aim, two main design approaches have been considered. The first one (adopted in the ITER CDA design) consists of copper stabilization loops (twin loops) attached to box-shaped blanket segments which are electrically and mechanically separated along the toroidal direction. In the second design approach (under consideration for the ITER EDA design), relying on a lower plasma elongation, no specific stabilization loops are required and the passive stabilization is achieved by toroidally continuous components, in particular by the plasma facing wall of the blanket segments, electrically connected along the toroidal direction, thus allowing a toroidal current to flow during the electromagnetic transients. In both cases electrodynamic loads arise in the blanket structures during plasma disruptions and/or vertical displacement events. A comparison between the two design approaches has been carried out from the eddy-current and related load distribution viewpoint.

Disruption problematics in segmented-blanket concepts

In tokamaks, the hostile operating environment originated by plasma disruption events requires that the first-wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence, there is a need to improve the safety features of the segmented-blanket design concepts in order to satisfy the disruption problematics. The present paper describes recent investigations on internal blanket reinforcement systems needed in order to improve the first-wall/blanket/shield structural design for next-step and commercial fusion reactors. Particularly in the context of SEAFP and ITER activities, representative 3-D CAD models of the inboard and outboard blanket regions and the related magnetomechanical simulations are illustrated.

Thermal-mechanical analyses of the ITER shielding blanket designs

Abstract Thermal-mechanical analyses were performed for the ITER shielding blanket design to define the performance of the structures when subjected to thermal loads, the internal pressure and the electromagnetic forces. The shielding blanket design under consideration has austenitic steel structure, low temperature and pressure water coolant, and a copper layer coated with beryllium for the first wall. The first wall is structurally integrated with the shielding blanket by thin lateral walls welded to the back plate. At the side wall of the sector, the internal pressure is sustained by the adjacent canisters. Two-dimensional analyses of the equatorial cross-section were carried out using generalized plane strain boundary conditions to simulate the out of plane structure behavior. The mechanical loads used in the analyses consist of an internal coolant pressure of 2 MPa and electromagnetic loads on the first wall due to plasma disruption events of 1 MPa for outboard sections. Two thermal loads were included: a surface heat flux of 50 W cm −2 on the first wall and nuclear heating from the neutronics analyses.

Modelling of transient electromagnetics in Tokamaks during off-normal conditions

During plasma disruption events in Tokamaks, a considerable amount of magnetic and thermal energy is associated to the transfer of plasma current into eddy and Halo currents. In this paper, a predictive numerical modelling is described concerning plasma-wall interactions during disruptive instabilities. Preliminary results are presented, giving an estimation of heat transfer interaction with electromagnetic phenomena. Eddy currents and heat deposition increase significantly with decreasing disruption time. An estimation of the order of magnitude of Halo currents and associated forces on plasma-facing conducting components is also presented.

Numerical simulation of vapor shielding and range shortening for ions impinging on a divertor during plasma disruptions

The behavior of divertor materials under the high heat load during a plasma disruption, a key problem in tokamak fusion technology, is studied at several laboratories under simulated plasma disruption conditions. Numerical modelling of the complex processes taking place at the divertor during a plasma disruption event was started recently at Kernforschungszentrum Karlsruhe. The aim of these activities is to determine the erosion rate of divertor material under different heat loads, to quantify the vapor shield effect and to give insight in the mechanisms of the beam target interaction. The simulation studies are performed with the Karlsruhe target code KATACO. This is a 1D plasma hydrodynamic code including beam energy deposition and multigroup radiation transport. First results for 10 and 100 keV proton beams with a power density of 5 and 50 MW/cm 2 respectively at a pulse duration of 100 μs show, that an equilibrium state at a plasma temperature of a few electron volt is reached shortly after the beginning of the disruption. A substantial part of the incoming energy is radiated back into the cavity.

Structural design aspects of the locking system needed for the next tokamak plasma-facing components against electromagnetic transients

A technical key issue during the design of the near-term Tokamak fusion devices (NET and ITER) is related to the structural problems of the needed attachments for the removable internal first-wall components (blanket segments and divertor plates) inside the vacuum vessel in the presence of major plasma disruption events. Both the understanding of the impact of the huge transient electromagnetic forces transmitted onto the vessel from the fastening/guiding support systems and the knowledge of the remote handling operations with related connection/disconnection of the first-wall component supports are involved. Among the various systems proposed as attaching locks, the “Twin-Belt” concept for the inboard blanket region is exposed. It has been developed at the TESLA Laboratory of the JRC-Ispra and has been retained by the NET/ITER Team as one of the options to be studied in detail. The use of such an intermediate toroidal, structurally continuous supporting system (belt bars) provides an independent system of attaching locks between the first-wall region and the vessel.

Instability analysis of first wall components of a tokamak reactor subjected to electromagnetic transients

The First Wall structure of a tokamak-type thermonuclear fusion reactor, such as NET (Next European Torus) or ITER (International Thermonuclear Experimental Reactor), should be able to withstand the combined effects of normal operating condition loads as well as significant electromagnetic transients induced during a plasma disruption event. Thus, stability analysis of such structures under large dynamic loads is essential for their adequate design. The stability analyses performed are based on incremental static analysis (RIKS method) on an imperfect structural geometry which is determined on the basis of the shape of the lowest buckling eigenvector obtained by a linearized buckling (EULER) analysis. Non linear behaviour due to large rotations and material behaviour, including strain-rate effects, were introduced in the incremental analysis.

3D eddy-current distribution in a tokamak first wall during a plasma disruption using “TRIFOU”

In fusion reactor studies there is a lack of knowledge concerning the electromagnetic-type of phenomena generated by a plasma disruption event (rapid quenching of the plasma current). The induced eddy current distribution in space and time in the passive conducting structural components surrounding the plasma ring needs to be accurately investigated. TRIFOU is a full 3D eddy-current computer program based on a mixed FEM and BIEM technique, using the magnetic field, h , as a state variable. It has already been used in various areas of interest including static or rotating machines, non-destructive testing, induction heating, and research devices such as tokamaks. It can take into account various geometries and a wide range of physical situations (time dependency, physical properties, etc.). The present application is related to the eddy-current situation arising from a strong electromagnetic transient generated in the NET (Next European Torus) first wall segment. With respect to previous numerical simulations, the general 3D approach for the current density shows different eddy current circulations in the front/side shells and in the stiff back plate. The results obtained by TRIFOU are illustrated by means of advanced computer graphic displays and an animation movie.

Structural design problems related to plasma disruption events of the removable net first wall segments

During a major plasma disruption event, the structural components of the NET (Next European Torus) tokamak reactor are subjected to strong electromagnetic transient loads arising from the interaction of the induced eddy currents with large magnetic fields. In the NET design, relatively many removable internal segments facing the plasma are assembled from the top inside of the vacuum vessel. This paper considers the NET reference design aspects for the fastening and guiding systems of the removable internal segments and illustrates the structural analysis of the first wall modules, including the behaviour of their supports during a major plasma disruption. In particular, for three outboard segments, a complete transient dynamic numerical simulation of the first wall structure will be described using a thin shell formulation.