Plasma Vertical Displacement Events Research Articles

Numerical assessment of thermal performance of W/Cu monoblock during steady-state operation and plasma vertical displacement events under ITER-like conditions

This paper employs a two-dimensional conduction model to assess the spatiotemporally thermal performance of ITER W/Cu monoblock under ITER-like conditions. The model considers the heat conduction processes including the evaporation, radiation and melting of both W armor and Cu heat sink materials under a heat load either during steady-state discharge or during plasma vertical displacement event (VDE). Moreover, the model employs coupling cooling water condition instead of the fixed cooling temperature at bottom in authors’ previous work. For the steady-state case, the heat flux distribution on the target tile, the input data to the model, was specifically evaluated by a particle-in-cell code to take into account its geometrical effect. For the VDE case, the heat flux distribution was simply assumed to be uniform on the tile surface. The simulated results obtained from the heat conduction model indicate that cooling water conditions have important influence on the thermal performance of W/Cu monoblock, especially for Cu heat sink materials during steady state. In the case of VDE, the baffle tile surface would undergo intense vaporization and melting, while whether Cu heat sink materials melted was determined by the combination of the exposed heat load and its duration. Based on typical VDE heat loads, detailed analysis is presented herein.PACS: 52.55.Fa, 52.55.Rk, 52.40.Hf, 52.65.-y

Vertical displacement events analysis using MAXFEA code in combination with ANSYS APDL in the final design stage of the DTT vacuum vessel

In the context of the Divertor Tokamak Test facility (DTT) project, a recently developed procedure based on the use of MAXFEA code in combination with ANSYS APDL already presented in [1] has been applied to study the EM and structural response of the DTT vacuum vessel (VV) during plasma Vertical Displacement Events (VDEs). From this study, it has emerged that the vacuum vessel mechanical response is acceptable in all of the scenarios but, during a particular event, namely the down-ward VDE slow scenario, a localized stress concentration in the lower side of the vacuum vessel has emerged. Downstream this analysis, mechanical reinforcements in this zone have been introduced in order to optimize the design. The aim of the paper is to show the successful application of this procedure in this context and its reliability as supporting tool in the final stage optimization design of these kind of components. The abovementioned down-ward VDE slow scenario is reported, showing a vertical force peak of around -10MN as the result of the combination of eddy and halo currents during the Current Quench (CQ). The structural analysis highlights that the EM loads occurring at this time instant produce a stress concentration localized in correspondence of the connection of the pedestal with the VV outer shell. This stress concentration has suggested to reinforce this zone in order to optimize the vacuum vessel design.

A Few Points of the Engineering Logic Discussed in ITER EDA on Evaluation of Halo- and AVDE-Induced Loads in Tokamaks

All tokamaks are designed to withstand a certain number of energetic electromagnetic (EM) transients caused by uncontrolled terminations of plasma pulses, including symmetric and asymmetric plasma vertical displacement events: VDEs and AVDEs. These events generate significant pulsed EM loads in all conductive components and coils. Axially symmetric transient EM loads induced by VDEs without Halo current have been calculated well since the 1980s; however, Halo-related EM load components and lateral loads associated with AVDEs still cause discussions. The author worked on fast plasma and EM transients in tokamaks quite a while ago then deviated to other areas but has been keeping track of the topic since. He is aware of discussions of the modelling of Halo currents and of significant scatter present in current estimates for AVDE-induced lateral loads and contends that some points of engineering logic formulated earlier on this topic may help reduce these uncertainties. This article summarises a few points of the engineering understanding developed in informal discussions within the ITER EDA team with the purpose to preserve these points for all tokamak developments.

Dynamic behaviour of DEMO vacuum vessel during plasma vertical displacement events

In tokamaks during plasma vertical displacement events (VDEs) electrical currents occur in the vacuum vessel (VV) and in the in-vessel components (IVCs) that interact with the strong magnetic field generating large electromagnetic (EM) loads. The aim of the present work is the evaluation of the peak displacements and accelerations of the DEMO VV in case of VDEs. Four types of VDEs are investigated that are specified to generate the most severe net loads. The assessment considers also the electromagnetic-mechanical coupling between the VV and the toroidal field (TF) coils whose characteristic stiffness and damping coefficients were determined through a transient EM analysis. In a second step a transient dynamic analysis was performed using a structural finite element (FE) model of the VV to evaluate the displacement and acceleration peaks during the VDEs in relevant locations of the VV.

Eddy Current Analysis of TF Coil Case for EAST

Plasma major disruptions and plasma vertical displacement events represent two extreme examples of fast transient electromagnetic processes, which will induce large current and electromagnetic loads that exceed those of loads during magnet normal operation and seismic loads. Eddy current and electromagnetic loads of the conducting structures need to be analyzed to verify the mechanical reliability of the tokamak during these transients. An integrated module, which consists of the mechanical design of the three-dimensional computer-aided design (CAD) solid modeling, generation of the finite element modeling, and external parameters input into the module, was integrated to perform the calculation of the electromagnetic loads and power losses. The developed integrated module can be used to calculate the Eddy current, electromagnetic forces, and power losses of the toroidal field (TF) coil-case for EAST tokamak during plasma disruptions and vertical displacement events. The simplified geometric model with 22.5° sector for EAST tokamak was used in view of the toroidal symmetry. The plasma current and position as functions of time are given by the numerical method. In this article, the module integration and evaluation of the electromagnetic loads and power losses on the conducting structures of the EAST tokamak by the finite element methods are described.

Electromagnetic-mechanical coupling method and stress evaluation of the Chinese Fusion Engineering Test Reactor helium cooled solid breeder under a vertical displacement event scenario

The helium cooled solid breeder (HCSB) blanket is one of the candidate blanket schemes proposed for the Chinese Fusion Engineering Test Reactor (CFETR). Plasma vertical displacement events (VDE) are considered to be important for evaluating the structural integrity of the blanket. The change of the magnetic field caused by the movement of the plasma and the rapid loss of current induces extra electromagnetic loads, which will cause significant structural damage to the blanket. This study aims to investigate the electromagnetic-mechanical coupling method and the coupling effects on the HCSB under plasma VDE. Based on the design of the HCSB for the CFETR, a finite element model is established by using the MAXWELL software, in which 69 current filaments are used to describe the plasma. The distribution of the induced current and magnetic field inside the HCSB under the plasma VDE, as well as the electromagnetic force are obtained by the ANSYS finite element software. Meanwhile, the distributions of stress inside each component of the blanket are obtained under electromagnetic-structure load and an evaluation of the stress is carried out. The results show that the existing design meets the requirements when plasma VDE occurs.

An FEM-BEM method for halo current problem and its application to HL-2M Tokamak

Halo current in Tokamak device during plasma vertical displacement events (VDEs) may damage the in-vessel components due to huge electromagnetic (EM) force. It is important to simulate the distribution of halo current in order to evaluate the EM force and the corresponding structural dynamic response. In this paper, an FEM-BEM method was proposed at first to simulate the current distribution in conductors excited by a sinusoidal conducting current injection. The response under a transient excitation was then realized by using the strategy of the frequency domain summation (FDS). A numerical simulation code was developed based on the proposed numerical scheme. The validity of the numerical method and corresponding code were proved by comparing numerical results of the developed code with those of the ANSYS software using step by step time domain integration. Finally, the proposed method was applied to the simulation of the halo current in a simplified vacuum vessel model of the HL-2M Tokamak. The simulation results provide good reference for the subsequent analysis of mechanical behaviors of in-vessel structures.

Development of load specifications for the design of ITER diagnostic system and port integration

ITER is the world’s largest fusion device currently under construction in South of France with 50+ diagnostic systems installed inside the port plugs, the interspace or the port cell region of various diagnostic ports. The plasma facing Diagnostic First Wall (DFW) and its supporting Diagnostic Shielding Modules (DSM) are designed to protect the front-end diagnostics from plasma neutron and radiation while providing apertures for diagnostic access to the plasma. The design of ITER port plug (PP) structure assembly including the DFWs, DSMs with in-port diagnostics is largely driven by the electromagnetic (EM) loads induced on these passive structural components during plasma disruptions, and the steady-state thermal loads from plasma volumetric and surface heating during normal operations. The design is significantly influenced by the dynamic response of in-port components attached to the port-specific DSM or PP closure plate as a result of transient EM loads induced on the PP assembly, vacuum vessel (VV) and the port extension during asymmetric plasma Vertical Displacement Events (VDEs). We investigate in this study the worst plasma disruption load cases for diagnostic systems of varying size and location, and summarize the steady-state thermal, transient EM and VV inertial loads for ITER diagnostic systems.

Electrical connectors for blanket modules in ITER

Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak.The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was developed in 2012 on the basis of more thoughtful analysis of bi-directional cyclic loading conditions influencing a fatigue lifetime. Detail comparative simulations of current and field patterns and subsequent analysis of the fatigue strength and technological assessment allowed make a final choice for the E-strap design in ITER.

Erosion and erosion products of tungsten and carbon-based materials irradiated by a high energy electron beam

A pulsed electron beam test was used to simulate the thermal response of plasma facing materials under high heat loads during plasma disruptions, particularly for relatively ‘long’ duration off-normal events such as plasma vertical displacement events. The erosion and erosion products of purified tungsten made by powder metallurgy (PM-W), W and B 4C plasma sprayed coatings on CFC or copper and C/C composite were investigated. The time evolution of the target current indicates that the erosion products or screening clouds partly or completely shield the incident electron beam, leading to a reduction of the target current and changes in the target current profile. The weight loss of PM-W is smaller than that of the tungsten coating or C/C composite, which is in agreement with the erosion behavior. Erosion products indicate that evaporation is a major erosion mechanism of PM-W, and exfoliation and particle emission are the dominant erosion mechanisms of carbon-based materials and W plasma sprayed coatings.

Effects of plasma disruption events on ITER first wall materials

In ITER, plasma disruption events may occur producing large fast thermal transients on plasma facing materials. Particularly important for the integrity of the first wall (FW) are relatively ‘long’ duration off-normal events such as plasma vertical displacement events (VDE) and runaway electrons (RE). An analytical methodology has been developed to specifically assess the effect of these events on FW plasma facing materials. For the typical energy densities and event duration expected for the primary and baffle FW, some melting and evaporation of the FW armor will occur without the beneficial effect of vapor shielding, and the metallic heat sink may also be damaged due to over-heating. The method is able to calculate the amount of melted and evaporated material, taking into account the evolution of the evaporated and melted layer and to evaluate possible effects of local temporary loss of cooling. The method has been used to analyze the effects of VDE and RE events for ITER, to study recent disruption simulation experiments and to benchmark experimental and analytical results.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part I: Theory and description of model capabilities

RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under ‘slow’ high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case.

The ITER Safety Approach for External Events

ITER will be the first large-scale tokamak to be designed as a nuclear facility to provide public protection from external hazards such as earthquakes. The design approach for such events has been developed consistent with ITER's moderate hazards and overall safety approach on a basis of the ITER site assumptions. Seismic design is described including selection of ground motions for design purposes, seismic safety requirements, and the seismic classification scheme. The results of preliminary seismic assessments are summarized including the potential for seismically induced plasma vertical displacement events (VDE). Finally, potential facility modifications available to deal with site-specific external hazards are suggested. At the Detailed Design Report stage of the Engineering Design Activity (EDA), it is concluded that ITER has been designed to deal with the site design assumptions for earthquakes and can be designed to safety cope with a range of site-specific external hazards with modest changes to the facility.