Plasma Disruption Events Research Articles

Investigation of intrinsic error fields in MAST-U device

In magnetic fusion devices, undesired non-axisymmetric magnetic field perturbations, typically called error fields (EF), have been observed to have a detrimental effect on plasma stability and confinement and can lead to brute plasma terminations, i.e. plasma disruption events. The main strategies that can be adopted to minimize the effect of EFs on the plasma dynamics consist in a careful alignment of the coils, when assembling the fusion device, and, most commonly, in the use of EF correction coils which counteract the non-axisymmetric fields by prescribing properly designed correction currents. In this work, an assessment of the n=1 EF source in MAST-U is presented. When constructing the MAST-U device, an optimization of poloidal and divertor coil positions has been adopted to reduce the n=1 EF source. This optimization consisted in the application of coil shifts and tilts, of the order of mms and mrads, respectively. To investigate the presence of a residual n=1 EF dedicated studies have been performed during the first MAST-U campaign. The compass scan method was employed to identify the n=1 EF, relying on the detection of the locked mode (LM) onset, which proved to be a challenging task. Therefore, a methodology based on Transfer Functions (TFs) among each coil and the n=1 radial magnetic field has been developed which allows the detection of LM formation. Such method is described here, complemented with the experimental results achieved, which suggest that the intrinsic n=1 EF source on MAST-U is relatively small with respect to MAST. Indeed, the empirical correction currents for n=1 EF minimization are smaller, about 0.2 kA, than the ones used in MAST, 1 kA range. This proves that the optimal coil alignment for n=1 EF minimization has been a successful strategy in MAST-U.

Electromagnetic analysis on conceptual design of the EU DEMO EC equatorial laucher

The EU DEMO Tokamak is foreseen to be equipped with Electron Cyclotron (EC) launching systems for plasma heating, MHD control and thermal instability suppression. Up to six launchers will be installed into equatorial ports with the aim to inject a maximum of 130 MW millimeter wave power at dedicated positions into the plasma. The integrated design of the EC launcher is made of two equatorial port plug modules mounted into the equatorial port extensions of the vacuum vessel that mainly serve as housing for the eight optical system´s mirrors.To guarantee reliable operation of the launcher, in the latest year, a structural system has been designed which provides secure fastening and alignment of the optical components, sufficient heat dissipation and protection of sensitive areas against nuclear loads from neutrons by adequate shielding capability.To prove the fastening concept of the port plug modules and mirrors, a set of mechanical loads acting on these components is essential. EM loads caused by plasma instabilities are major contributions of mechanical load set for components close to the plasma.The present paper describes the EM analysis carried out to support the assessment of the pre-conceptual design solution. A dedicated Finite Element (FE) model was developed using ANSYS® solid236 element type (edge-based formation). Transient electromagnetic analysis was performed under centered plasma disruption event. Total force and moment due to Lorentz and ferromagnetic loads on the EC launcher port plugs and mirrors were computed. The resulting EM loads can be used for the preliminary design assessment.

Preliminary electromagnetic analysis of the COOL blanket for CFETR

The supercritical CO2 cOoled Lithium-Lead (COOL) blanket has been designed as one advanced blanket candidate for the Chinese Fusion Engineering Test Reactor (CFETR). This work focuses on the electromagnetic (EM) loads (Maxwell force and Lorentz force) acting on the COOL blanket, which are important mechanical loads in further structural analysis of the COOL blanket. A 3D electromagnetic analysis is performed using the ANSYS finite element method to obtain EM loads on the COOL blanket in this study. At first, the magnetic scalar potential (MSP) method is used to obtain the magnetic field and the Maxwell force on the COOL blanket. Then, the magnetic vector potential (MVP) method is performed during a plasma disruption event to get the eddy current distribution. At last, a multi-step method is adopted for the calculation of the Lorentz force and the torque. The maximum Lorentz forces of inboard and outboard blanket structural components are 5624 kN and 2360 kN respectively.

DECAF perks up plasma forecasting

Physics-based approach accurately predicts plasma disruption events in tokamaks producing fusion energy

Electromagnetic-thermal two-way coupling analysis and application on helium-cooled solid blanket

The blanket plays an important role in fusion reactor and stands extremely high thermal and electromagnetic loads during operation situation and plasma disruption event, brings the need for precise thermal and electromagnetic analysis. Since the thermal field and EM field interact with each other nonlinearly, we develop a method of electromagnetic-thermal two-way coupling by using finite element software COMSOL. The coupling analyses of blanket under steady state and MD event are implemented and the results are analyzed. For steady state, the influences of coupling effects are relatively small but still recommended to be considered for a high precision analysis. The influence of thermal field on EM field can't be ignored under MD events. The variation of force density could cause a significant change in stress in certain parts of blanket. The influence of Joule heat during MD event is negligible, yet the potential temperature rise caused by induced current after MD event still needs to be researched.

Preliminary thermal optimization and investigation of the overall structural behaviour of the EU-DEMO water-cooled lead lithium left outboard blanket segment

The conceptual design phase of the EU-DEMO reactor has been recently launched, with the aim of evolving the DEMO pre-conceptual layout towards a more robust and articulated geometric configuration able to cope with most of the design requirements and to show further margins for the passing of the current potential show-stoppers. Hence, the achievement of the conceptual design of the Water-Cooled Lead Lithium Breeding Blanket (WCLL BB) is one of the milestones the EUROfusion consortium aims to achieve in the close future. To this purpose, within the framework of the research activities promoted by EUROfusion, a research campaign has been launched at the University of Palermo, in close cooperation with ENEA-Brasimone, in order to investigate, from the thermal and thermomechanical standpoints, the WCLL BB Left Outboard Blanket (LOB) segment performances. The scope of the analysis carried out has been the advancement of the WCLL LOB segment design in view of the prescribed thermal and structural design requirements. To this end, the nominal WCLL BB operating conditions as well as accidental steady-state scenarios derived from the in-box loss of coolant accident and the upper vertical plasma disruption event have been considered. In the first part of the study, the internal Double-Walled Tubes (DWTs) layout of the WCLL LOB equatorial region have been preliminarily optimized so to find a geometric configuration able to cool the segment structure maintaining the maximum temperature below the suggested limit of 550 °C. Then, a purposely set-up analytical procedure has allowed obtaining a set of interpolating functions able to accurately reproduce the calculated thermal field and to extrapolate it to the whole WCLL LOB segment. Afterwards, the thermo-mechanical analysis of the entire segment has been performed under the selected loading scenarios. The structural behaviour of the segment has been assessed in compliance with the RCC-MRx code, adopting the set of criteria on the basis of the nature of the considered loading scenario. The obtained results, herewith presented and critically discussed, have allowed introducing significant design improvements with respect to the WCLL LOB segment reference geometric layout, coming from the pre-conceptual design phase, paving the way for future, and more detailed, structural investigations.

Analysis of the Thermo-Mechanical Behaviour of the EU DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment under an Optimized Thermal Field

Within the framework of the EUROfusion research activities on the DEMO Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) design, a research study was performed to preliminarily optimize, from the thermal point of view, the WCLL Central Outboard Blanket (COB) segment in order to investigate its structural behaviour under a realistic thermal field. In particular, a study of thermal analyses was performed to optimize the Double Walled Tubes and Segment Box cooling channels’ geometric configurations along the poloidal extension of the WCLL COB segment, in order to obtain a spatial temperature distribution fulfilling the thermal design requirement. Then, the thermo-mechanical analysis of the WCLL COB segment under Normal Operation (NO, representing nominal conditions), Upper Vertical Displacements Event (UVDE, representing a plasma disruption event) and Over-Pressurization (OP, representing an in-box loss of coolant accident) scenarios were carried out, assuming the previously obtained thermal field, to realistically predict displacement and stress fields. Finally, a stress linearization procedure allowed comparing the stress values obtained in some critical regions of the structure with the criteria prescribed by the reference design standard RCC-MRx. A theoretical–numerical approach based on the Finite Element Method (FEM) was followed using the commercial code Abaqus v. 6.14.

3D-FOX—A 3D Transient Electromagnetic Code for Eddy Currents Computation in Superconducting Magnet Structures: DTT TF Fast Current Discharge Analysis

Bulky metallic structures are needed in the toroidal field (TF) superconducting magnets for fusion applications to withstand the large Lorentz forces acting on the winding. The pulsed coil operation during a plasma scenario and the fast current discharge, or a plasma disruption event in off-normal operating conditions, cause transient magnetic fields, inducing eddy currents in the TF structures. The eddy currents generate heat in the structures heating in turn the winding pack, eroding the temperature margin: such power deposition is a key input for thermal-hydraulic (TH) analyses. However, the computation of eddy currents in fusion magnets is a challenging topic since a transient, fully 3D electromagnetic (EM) model is required. The EM problem is solved here by means of the finite element (FE) open source code FreeFEM++. First, the correct implementation of the EM problem is verified by means of suitable benchmarks against both simple analytical cases and the results obtained with state-of-the-art FE commercial codes on the DTT TF coil, used as a reference geometry. Then the EM code is applied to the evaluation of the magnetic fields and eddy currents induced in the same reference coil during the normal (static) and off-normal (transient) operation; the output of the EM analysis is used as input to the TH analysis carried out with the 4C code, aimed at computing the temperature margin evolution during the transient.

Analytical solution of tokamak vibrations during axisymmetric plasma disruptions

Tokamaks are structures containing a massive supporting structure housing a vacuum vessel with hot plasma. Due to the vacuum vessel heating and temperatures up to 500 ⁰C as for the COMPASS-U, flexible connections need to be used between the main support and the vacuum vessel to accommodate the thermal expansion. Large electromagnetic forces are generated during plasma disruption events, such as Vertical Displacement Events (VDE), inside the vacuum vessel and in the coil system: Central Solenoid (CS), Poloidal Field (PF) coils and Toroidal Field (TF) coils directly attached to the support. In the present paper a parametric analytical model of tokamak vibrations due to axisymmetric plasma disruptions was presented, allowing predictions of the vacuum vessels and the main tokamak structure displacements, velocities and accelerations as well as the dynamic forces transmitted to the foundations. A tokamak was modelled as a 2 DOF dynamic system, including the mass of the main support and the vacuum vessel and the axial stiffnesses and damping values of their supports. A close-form analytical solution was found assuming proportional (Rayleigh) damping for the most general shape of the force functions – described with a series of piecewise linear functions. The electromagnetic forces during axisymmetric VDE were derived from a dedicated 3D model. Detailed solution for the most recent design of the COMPASS-U tokamak was presented, as well as a number of parametric studies checking the sensitivity to the design parameters. The conditions at which large part of the electromagnetic force is transmitted to the concrete foundations were identified. The accuracy of the derived analytical solution was verified with direct numerical integration of the equations of motion.

Numerical simulation on the eddy current electromagnetic forces and the mechanical reliability of ITER in-wall shielding blocks induced by plasma disruption events

In-wall-shielding (IWS) blocks located in ITER vacuum vessels (VVs) are used for neutron shielding. They are subjected to complex eddy current electromagnetic forces during plasma disruption events, especially for IWSs of irregular sectors. Reliable evaluation of this kind of electromagnetic force is of great significance for the structural design of IWSs as well as the safety of fusion reactors. In this paper, the eddy current electromagnetic forces on 121 IWS blocks in the irregular sector and their associated conductive parts were numerically calculated based on one of the most important disruption events (the MD UP LIN36ms disruption event). To promote the accuracy of the simulations, the excitation loads during the plasma disruption current were processed from the database simulated by DINA code into the FE (finite element) model of plasma with a special transformation procedure. The results show that the method of applying plasma current from DINA to the finite element model is reasonable and accurate. The electromagnetic forces of the blocks are demonstrated to be reliable. The block with the maximum electromagnetic force is selected for mechanical analysis, and the reliability of its components is verified.

Using MAXFEA code in combination with ANSYS APDL for the simulation of plasma disruption events on EU DEMO

Plasma 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.

Design and verification status of COMPASS-U vacuum vessel and stabilization loop

The COMPASS-U is a new tokamak in the design phase at Institute of Plasma Physics, Prague, Czech Republic. It is a medium-sized tokamak with high magnetic field (Bt = 5 T) and high temperature vessel (< 500 °C). Hence, large electromagnetic forces are expected on the Passive Stabilizing Plates (PSP) and Vacuum Vessel (VV) during plasma disruption events. A non-selfconsistent approach is selected for the electromagnetic calculations using ANSYS Maxwell software to determine the highest probable forces on passive structures and to design a robust vacuum vessel.This paper outlines the latest status on the design activities of the vacuum vessel concept and its mechanical verification against the electromagnetic forces. The vessel design concept is developed considering the spatial constraints, fabrication simplicity, minimum cost and the design is verified for minimum deformation and stresses in the vessel. It also includes the design evolution of PSP, PSP support and VV support. The mechanical analyses are carried out for load combinations of plasma transient events.

Electromagnetic model for finite element analyses of plasma disruption events used in the design phase of the COMPASS-U tokamak

Large electromagnetic forces acting on the vacuum vessel and passive stabilizing plates of the COMPASS-U tokamak (R = 0.894 m, a = 0.27 m, BT ≤ 5 T, IP ≤ 2 MA) due to disruption events are expected during its operation. An electromagnetic model based on the finite element method was developed using ANSYS software to determine the maximal possible forces on the vacuum vessel assembly that might occur. The effects of current quench, thermal quench, vertical displacement events and halo currents are all incorporated in the model. The toroidal eddy currents, the poloidal eddy currents (caused by thermal and current quench) and the poloidal halo current are taken into account. The model predicts that the induced toroidal current can reach the value up to 1.33 MA in the vacuum vessel shell and the value up to 0.52 MA in the passive stabilizing plates. Vertical force up to 4.1 MN acting on the entire vacuum vessel assembly can be expected and force up to 2 MN acting on the stabilizing plates, in particular. The results of the presented electromagnetic model are part of the load specification for the mechanical design of the vacuum vessel assembly of the COMPASS-U tokamak.

Effect of ITER CS and PF magnets on EM loads outside vacuum vessel at plasma disruption events

One of the most crucial issues in the design of the ITER machine is the electromagnetic (EM) loads associated with eddy currents induced in the conducting structures during plasma disruptions. The ITER database contains tens of calculation scenarios for possible plasma disruption events. The duration of these scenarios is usually limited by the end of the plasma current quench when the toroidal plasma current decreases to zero. For further EM analysis, the currents in the poloidal field coils are conventionally assumed constant after the end of the current quench. Such approach is not good for estimation of electromagnetic loads acting on the tokamak components located outside vacuum vessel. A possible way to solve this problem is to employ pre-determined coil current variations over the entire period of observation including time interval after the current quench. This paper describes an approach to such electromagnetic calculations and comparative results for selected cases with/out an extended current-time specification.

A Simplified Analytical Model for the Analysis of Magnetomechanical Dynamic Response of a Test Module in J-TEXT Tokamak

Test blanket modules (TBMs) are key components of the ITER for the demonstration of tritium breeding. During plasma disruption, the TBMs and other in-vessel components suffer large electromagnetic forces and vibrations accompanied by a significant magnetostructural coupling effect. Aiming to evaluate the coupled vibration of a test module in J-TEXT Tokamak, a simplified analytical model was developed in this paper. The model has one degree of torsional freedom and was established by the idealization of the geometry and the confining magnetic field. The analytical solution of the model indicates that the magnetic stiffness effect is not significant for this problem, and the model gives a conservation estimation of the magnetic damping effect. To demonstrate the feasibility of the analytical model, the dynamic response of the test module in J-TEXT during a typical plasma disruption event was simulated by using an FEM numerical code of the Lagrangian sequential coupling method and was found to be in good agreement with the analytical solution. To prove the validity of the numerical model and code, the coupled vibration of a cubic block in a large constant magnetic field and a pulsed magnetic field of an electromagnet was measured by fabricating an experimental setup in a large magnet and was also calculated by using the FEM code under the experimental conditions. The measured results agree well with the numerical ones, which proved the validity of the analytical model indirectly. Due to the explicit form of the coupling term, the analytical solution gives a good reference when considering the magnetostructural coupling effect of a typical in-vessel structure in the tokamak device.

ITER Torus Diamond Window Unit: FEM Analyses and Impact on the Design

The ITER torus diamond window unit is part of the electron cyclotron (EC) upper launcher (UL) used to direct high power microwave beams generated by the gyrotrons into the plasma for heating and current drive (H&CD) applications. The UL consists of an assembly of ex-vessel waveguides (WGs) and an in-vessel port plug (PP). The diamond window units form vacuum and confinement boundaries between the torus volume and the transmission lines (TLs) which guide beams between 1 and 1.5 MW at 170 GHz from the gyrotrons to the launcher. There are eight window units attached to the assembly of the WGs, one unit for each WG. The design strategy of the unit is to have a very rigid outer frame able to withstand the potential external loads acting on the unit while thin copper cuffs brazed to the diamond disk allow indirect cooling of the disk to remove the EC power absorption during the beam transmission. The load combination given by the stringent ITER seismic level 2 (SL-2) event occurring during baking of the torus vessel is the design driver for the outer frame of the unit. In fact, the assembly of the WGs is connected from one side to the ceiling of the ITER port cell area by a support frame and to the UL PP from the other side. Movements of the torus vessel due to baking, seismic and plasma disruption events, result in forces and moments acting on the units. Furthermore, during a seismic event, the unit is subject to additional loads induced by the oscillation of the support frame attached to the ceiling. An outer frame surrounding the window unit is thus required to ensure the structural integrity and the confinement function of the unit. This paper shows how the design of the window unit was assessed and optimized by finite element method (FEM) analyses. A specific methodology was developed to carry out the analyses with respect to the seismic and baking loads. At Karlsruhe Institute of Technology (KIT), FABRY-PEROT resonators measure the loss tangent of the diamond disk and it is then used as input to the computational fluid dynamics and FEM analyses aiming to assess the design. The impact of the FEM analyses on the design of the window unit is discussed also together with the manufacturing aspects of the unit.

ITER magnetic sensor platform engineering analyses

ITER in-vessel magnetic sensors play a key role for ITER plasma operation. Each sensor is accommodated in a platform mounted on the inner surface of ITER vacuum vessel and behind the blankets.A full set of engineering analyses has been performed on the platform to assess the feasibility of the design configuration.Electromagnetic Sub-Modelling technique has been used for very detailed evaluation of the EM loads due to 8 category III plasma disruption events acting on platforms at 23 locations from a full poloidal array. The two cases yielding highest EM load were selected to complete the structural assessment.Thermal analysis was performed for a platform located where maximum heat load is expected and the structural assessment was performed considering a pessimistic input load combination of highest EM loads and highest thermal stresses over ITER operation lifetime.An optimization process to reduce the EM loads and increase the structural strength of the platform was set up and it led to positive results on the structural integrity of the platform structure with the optimized design.Future work is planned to incorporate the thermal deformation of VV surface and refine the calculation.

Mechanical design of the high-powered helicon antenna and stripline feed in the DIII-D Tokamak

A high-powered “comb-line” helicon antenna for use within the DIII-D Tokamak is currently in design and fabrication at General Atomics. The antenna will drive current in high beta discharges using electromagnetic helicon waves. The high-powered helicon antenna (HPHA) is expected to couple up to one MW of power into DIII-D plasmas at a frequency of 476 MHz. The antenna design includes 30 individual, inductively coupled modules fastened to six internally water-cooled Inconel back-plates. The end modules have special design features to provide RF stripline feed attachment points. Each back-plate is mounted to a pedestal, which is secured to the vessel wall via a combination of studs and welded brackets. The mechanical design includes features to minimize and survive disruption forces, thermal loading from RF losses, vacuum vessel bake (350C), and installation considerations. Thermal stress design challenges include plasma heat loads, thermal ratcheting, and bake-out cycles that result in thermal growth differences between the stripline and the vessel. Splitting the back-plate into six sections each with a central pedestal support significantly reduces the disruption loads. A single stripline folded in half, fed near the fold provides a 180° phase shift between the stripline module connections for driving the two module straps. This looped mono RF coaxial stripline design was chosen to enable using one feed-through, to optimize space limitations, ease of installation, structural rigidity, and RF tuning ability. The striplines are supported by a quarter-wavelength “stub”, mid span, for support during thermal cycling and plasma disruption events. Multiphysics FEA analyses are performed to optimize the geometric shapes to meet the aforementioned design challenges. A description of the mechanical design, fabrication, installation, and analyses of the HPHA and stripline feeds will be presented

A high-power helicon antenna for the DIII-D tokamak and its electromagnetic aspects

A comb-line antenna to demonstrate efficient off-axis non-inductive current drive from the absorption of toroidally directed very high harmonic fast waves is being designed and built for DIII-D. The antenna consists of a toroidal array of 30 modules, spanning 1.7 m on the outer vessel wall just above the tokamak midplane. This antenna will be fed with 1 MW of RF power at 476 MHz through a stripline (SL) feed on both sides inside the vacuum vessel. COMSOL Multiphysics, a commercial finite element analysis software, was used to perform the RF analysis and the induced force analysis due to plasma disruption events on the whole antenna system. Results on the RF performance of the SL and the RF coupling into and between the modules, and the SL and modules’ RF losses will be presented. In addition, the disruption induced current and forces are shown, and a method to mitigate them is presented.

Simulation of the electromagnetic wall response during Vertical Displacement Events (VDE) in ITER tokamak

The key basis for tokamak plasma disruption modeling is to understand how currents flow to the plasma facing surfaces during plasma disruption events. In ITER tokamak, the occurrence of a limited number of major disruptions will definitively damage the chamber with no possibility to restore the device. In the current exchange plasma-wall-plasma, according to the Helmholtz decomposition theorem, our surface current density in the conducting shell - the unknown of our problem - being a vector field twice continuously differentiable in 3D, has been splited into two components: an irrotational (curl-free) vector field and a solenoidal (divergence-free) vector field. Developing a weak formulation form and minimizing the correspondent energy functionals in a Finite Element approach, we have obtained the space and time distribution of the surface currents. We verified successfully our numerical simulation with an analytical solution with pure homogeneous Neumann B.C. and satisfying the necessary existence condition. By considering the iron core presence in JET tokamak, we have split the magnetization currents - the unknowns in some integral equations - into two components, the first producing a magnetic field in the iron region only and the second producing a magnetic field in the vacuum, obtaining thus a better evaluation of the influence of the iron core on the plasma equilibrium. To reduce the influence of the singularities appearing during the surface currents determination in multiply connected domains (L-shaped domains) we have used a conformal transformation method.