Vertical Displacement Events Research Articles

Preliminary electromagnetic loads calculation for the divertor of K-DEMO

The variation of plasma current and magnetic fields generated by superconducting magnet coils causes electromagnetic (EM) loads especially during the abrupt plasma current changes such as major disruption (MD) and the vertical displacement event (VDE) of plasma. The EM loads are one of the most important external loads for in-vessel components like blanket and divertor modules. The aim of this study is to estimate the EM loads on the K-DEMO divertor module under MD by using the commercial finite element software, ANSYS Mechanical/Emag. The superconducting magnet system of K-DEMO includes 16 toroidal field (TF) coils, 8 central solenoid (CS) coils, and 12 poloidal field (PF) coils. The magnetic field at the plasma center is about 7.4 T and the peak field is as high as ˜16 T. The conceptual model of the K-DEMO divertor module including outboard and inboard targets, dome, and the cassette body with connecting supports was developed. The 22.5° sector model applying symmetric boundary condition on cutting surfaces was employed to calculate EM loads efficiently. The reduced activation ferritic martensitic (RAFM) steel has been considered as the structural material of in-vessel components. Based on the current design concept of the K-DEMO, EM loads and eddy current are calculated.

Model development and electromagnetic analysis of a vertical displacement event for the Chinese Fusion Engineering Test Reactor helium cooled solid blanket

As one of the typical blanket concepts for the Chinese Fusion Engineering Test Reactor, a conceptual structure of a helium cooled solid breeder (HCSB) blanket was designed by the University of Science and Technology of China. Considering that electromagnetic load is one of the main concerns for the blanket module, a finite element method model of the HCSB is developed and the electromagnetic analysis of the blanket module is implemented using a finite element analysis software called MAXWELL. For transient electromagnetic analysis with the prescribed vertical displacement event, a more accurate model where the plasma is described by 69 filaments is adopted and the whole 15 blankets lying in a toroidal–radial section are explored. Research into the ferromagnetic effect of reduced activation ferritic/martensitic steel is carried out and the magnetic field, induced eddy currents, and magnetic force are computed and analyzed. The analysis results show that the ferromagnetic effect broadened the range of the magnetic field of the model and strengthened the eddy current effects. In addition, the maximum value of the eddy current density is 71.2 MA m−2 and the maximum magnitude of the electromagnetic forces is 1409.0 kN under the ferromagnetic effect.

Designs and technologies for plasma-facing wall protection in EU DEMO

The EU DEMO plasma is almost completely enveloped by large breeding blanket segments for tritium breeding and power extraction. Shaping of the blanket plasma-facing wall in 3D may prove to be essential, but this strategy alone is not sufficient to protect against anticipated transient plasma events. The high heat flux wall-limiter approach used in ITER is not thought to be viable in a tritium self-sufficient power reactor, and so in EU DEMO wall protection using discrete limiters is pursued. Two types of discrete limiter are described in this paper. One is an equatorial port limiter designed to handle the power during the plasma start-up phase, making use of water-cooled tungsten/CuCrZr monoblock technology. The second is the upper limiter, featuring a plasma-facing component designed specifically for extreme transient loading due to a vertical displacement event. The heat flux channelling and thermal barrier features of this design are shown to considerably reduce CuCrZr pipe temperature, and so reduce the likelihood of catastrophic failure. A preliminary neutronic calculation has shown that the impact of these discrete limiters on overall tritium breeding ratio is relatively low.

Structural assessment of HL-2M in-vessel cryopump under EM loads

HL-2M in-vessel cryopump was installed in the vacuum vessel, and was one of the important pumping systems during plasma operation. Electromagnetic (EM) fast transients would occur in HL-2M device, and the consequent loads should be considered for the design of the HL-2M cyropump, Such as major disruptions (MD), vertical displacement events (VDE), etc. In order to prevent the damage of the HL-2M cryopump structure under the EM loads, the structure analysis was performed based on the ANSYS software. The results show that the EM force of HL-2M cryopump induced by the plasma disruptions with a current quench time 4ms is smaller than that by VDE. However, all the above EM loads results could satisfy the structural design requirements for the HL-2M cyropump according to the allowable stress criteria of ASME Code. Meanwhile, the HL-2M cryopump could withstand the aforementioned EM loads with a cycle number of 1000.

Hot VDE investigation of the negative triangularity plasmas based on HL-2M tokamak

Heating loads on the divertor has been the major issue for the tokamak operation. The innovative negative triangularity plasmas are considered to be of the favorable property for the power handling. Investigating the hot-plasma vertical displacement events of the negative triangularity plasma has an important significance for the future fusion reactors. In this paper, the hot-plasma vertical displacement event of the negative triangularity plasmas are investigated by the DINA code, based on the new medium-sized copper conducting tokamak HL-2M. Simulation results show that: (i) The initial vertical instability, of the negative triangularity configuration, develops much more faster than that of the positive triangularity configurations; (ii) During the hot VDE, larger absolute value of the negative triangularity causes speeder development of the initial vertical instability and severer mechanical loads; (iii) During the hot VDE of the negative triangularity plasma, the most non-uniform distribution of the electromagnetic force almost occurs in the outer part of the vacuum vessel. Therefore, an arc design of this part is recommended; (iv) The initial vertical instability, as well as the electromagnetic loads, can be mitigated, by further optimizing the magnetic structure around the divertor region.

Optimization of DEMO geometry and disruption location prediction

The design study of DEMO is one of the main points of the European Roadmap to Fusion Electricity. The present pre-conceptual design phase of DEMO is used to explore a flexible range of the main machine geometrical design parameters and plasma scenarios. This paper presents the electromagnetic (EM) modelling of the DEMO baseline scenario, including the optimization activities on the electrically conductive structures surrounding the plasma and their influence on the passive vertical stabilization (VS). The improvements on the passive and active VS allowed increasing the maximum controllable plasma elongation, with a consequent increase on the fusion performance. Finally, a study is presented on the possibility to predict the plasma final position following a vertical displacement event (VDE). This prediction capability supports the development of a wall protection strategy from plasma transients. A model is presented based on the assumptions on the characteristic time constants expected for the current quench (CQ) time and L/R time constant of the conductive structures.

Effects of asymmetries in computations of forced vertical displacement events

Visco-resistive magnetohydrodynamic (MHD) computations with the NIMROD code (Sovinec C R et al 2004 J. Comput. Phys. 195 355) are applied to a model tokamak configuration that is subjected to induced vertical displacement. The modeling includes anisotropic thermal conduction within an evolving magnetic topology, and parameters separate the Alfvénic, resistive-wall, and plasma-resistive timescales. Contact with the wall leads to increasingly pervasive kink and tearing dynamics. The computed 3D evolution reproduces distinct thermal-quench and current-quench timescales, a positive bump in plasma current, and net horizontal forcing on the resistive wall. The MHD dynamo effect electric field, is analyzed for understanding the nonlinear effects of the fluctuations on the spreading of parallel current density and the resulting bump in plasma current. Forces on the resistive wall are consistent with Pustovitov’s analysis (Pustovitov V D 2015 Nucl. Fusion 55 113032); the plasma remains in approximate force-balance with the wall, so net force is accurately computed from integrating stress over the wall’s outer surface. Improvements to the modeling that are needed for predictive simulation of asymmetric vertical displacement events are discussed.

Intrinsic dust transport in ASDEX upgrade studied by fast imaging

Fast video data recorded from 2008 to 2012 in full-tungsten ASDEX Upgrade have been analyzed with the TRACE algorithm developed to automatically detect and track dust particles. The purpose of the work presented in this paper is to complete earlier study of the influence of various discharge conditions on dust rates with an investigation of dust motion under most remarkable conditions. The 3D trajectories of intrinsic dust particles have been confronted to simulations carried out with the DUCAD (Dust Characterization And Dynamics) code, based on the Orbital Motion Limited (OML) approach for dust/plasma interaction modelling. The motion of micrometric spherical dust grains is found to be largely dominated by inertia in the considered experimental conditions. The influence of Vertical Displacement Events (VDEs) and of various heating scenarios on dust areas of formation and trajectories is investigated. It is found that VDEs inject dust originating from the PFCs which are heated by contact with the plasma, upward VDEs having a stronger impact on dust rates and transport than downward VDEs. The use of Neutral Beam Injection (NBI) significantly enhances dust production and transport coming from sectors where injectors are installed. On the contrary, Ion Cyclotron Resonance Heating significantly reduces dust rates and transport in the whole plasma volume.

Deuterium retention in re-solidified tungsten and beryllium

Leading edges of the ITER tungsten (W) divertor are expected to melt due to transient heat loads from edge localized modes (ELMs), and melting of the entire divertor surface will occur during vertical displacement events (VDEs) and disruptions. In addition, understanding tritium retention in plasma facing materials is critical for the successful operation of ITER due to safety reasons. Thus, the question of how melting affects hydrogenic retention is highly relevant for fusion devices. Here we use an Nd:YAG laser to melt tungsten and beryllium in vacuo, and the samples are subsequently exposed to deuterium plasma with sample temperatures ranging from 370 to 750 K. The deuterium content in re-solidified and reference (no laser) samples is measured using thermal desorption spectroscopy and modeled using TMAP-7. In all cases, the re-solidified samples have lower retention compared to the reference samples. For re-solidified tungsten, the most significant effect is in the 1.8 eV trap with peak thermal desorption temperature of ∼750 K, which had a 77% reduction in the peak release rate compared with the reference sample. SEM imaging indicates that laser melting and re-solidification of tungsten anneals intrinsic defects that act as nucleation sites for larger-scale defects that develop during plasma exposure. However, melting does not significantly affect traps with lower de-trapping energies of 1.0 eV and 1.4 eV. In beryllium, melting and cracking results in lower retention compared to the reference sample by 40%, and thermal desorption profiles indicate that the diffusion depth of deuterium into re-solidified beryllium is lower than that of the reference sample.

Study of electromagnetic disruption forces for plasma detachment measurements in DEMO

Global plasma instabilities such as plasma disruptions and vertical displacement events (VDEs) can generate huge transient electrical currents in the conductive parts of the divertor. The interaction of the currents with the magnetic field induces Lorentz force loads. Particularly, the force induced by a VDE can impose a critical impact on the divertor water-cooling pipes due to high current density. A recently proposed isolated-target design solution for the DEMO divertor introduces a shunt element between the target plate and the cassette body to divert electrical current towards more massive cassette body. Such a design solution is also considered as a diagnostic tool for measuring thermo currents (voltages) through a shunt resistor to control plasma detachment.In this study, a finite-element analysis is performed to provide a conservative upper value for the shunt resistance by investigating the impact of the electromagnetic disruption forces on the integrity of water-cooling pipes in the isolated-target design solution of the DEMO divertor. In the analysis, a maximum allowed pipe current is estimated assuming realistic material properties and magnetic fields, and true geometries of the divertor cassette and water-cooling pipes fixed to the cassette. The results show that local pipe yielding (representing damage initiation) develops for electrical currents larger than ∼3.5 kA, which sets a conservative (assuming extreme VDE halo currents) upper value for the shunt resistance to ∼40 μΩ.

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.

Intrinsic dust dynamics and temporal correlation with plasma parameters in experimental advanced superconducting tokamak

The intrinsic dust dynamics for different discharge conditions and temporal correlation of dust events with plasma parameters have been first studied using a fast camera in Experimental Advanced Superconducting Tokamak. Dissimilarities in dust dynamics have been observed with dust coming out from upper tungsten and lower graphite divertors. Specific features like dust production at hotspot and dust ablation have been reported. It has been observed that in the ion cyclotron range of frequencies heating pulse duration dust events were significantly lower in both disruptive and non-disruptive discharges with combination of other auxiliary heating powers. The temporal correlation of dust events with plasma parameters in disruptive discharges preceded by vertical displacement events revealed the causality link between the dust release and disruptions. The detailed analysis of the dust rate in disruptive discharges has been presented.

Force-free motion of a cold plasma during the current quench

Cold disruptive plasma tends to move during the current quench. Its motion is essentially force-free since the current quench timescale is resistive rather than Alfvénic. In contrast to the hot vertical displacement events, the frozen-in condition is violated in the cold plasma case, and the plasma motion is not governed by magnetic flux conservation but rather by its dissipation. We present a numerical model of the cold plasma dynamics. This model predicts electromagnetic loads on the vacuum vessel, the plasma flow and density evolution, and the plasma centroid evolution. Our calculations include poloidal wall currents. We demonstrate their significant contribution to the force acting on the vacuum vessel.

Structural integrity analysis for manufacturing design of the ITER vacuum vessel gravity support

Due to the complicated operation condition and environment, the ITER Vacuum Vessel Gravity Support (VVGS) shall sustain the design loads including gravity of vacuum vessel and in-vessel components, thermal load, electro-magnetic loads from vertical displacement event and plasma disruption, and seismic load. Meanwhile, its dual hinges will allow radial movement of the VV caused by thermal expansion from room temperature to the operation (100 °C) and baking (200 °C) temperatures. This paper introduces structural integrity analysis of the manufacturing design of the ITER VVGS. Firstly, an overview of design description is followed by the new mechanical loading conditions which are specified and categorized with appropriate structural design criteria and analysis methodology. Secondly, protection against plastic collapse (P-type) and failure from cyclic loading (S-type) has been verified by elastic analysis or elastic-perfectly plastic analysis. Finally, the global structural integrity of VVGS against P-type and S-type damage has been justified with regards to RCC-MR according to the new loading conditions with a quantitative conclusion.

Upgrade of Cooling Channels for Four-Strap ICRF Antenna of EAST

In order to meet the heating requirement of EAST tokamak, ion cyclotron range of frequency (ICRF) heating is utilized as one of the main auxiliary heating methods, which plays an important role in coupling radio frequency (RF) power into the plasma. As a member of plasma facing components, ICRF antenna suffers the plasma radiation as well which is one of the causes of the thermal load on the antenna. Plasma radiation and RF thermal load during the operation of ICRF and thermal load caused by HALO current during vertical displacement event will contribute together to a huge thermal load on the surface of the antenna and consequently the structural stability and reliability of ICRF antenna will be affected. Therefore, cooling channels is needed to remove heat on the ICRF antenna. In this paper, cooling channels for straps of ICRF antenna are designed based on the thermal load on the antenna, and thermal-structural analysis of straps with cooling channels is also carried out by the finite-element method.

Power handling limit of liquid lithium divertor targets

A model is formulated to make a first estimate of the maximum tolerable power of liquid lithium (Li) divertor targets, and to gain insight into their behaviour in terms of Li loss rate and surface temperature. The model, formulated as a simple analytical expression, states that the incoming power is balanced by heat conduction through the target and by the Li which dissipates energy via evaporation, radiation and ion-neutral friction. A target is considered to fail when the net Li loss flux from the surface exceeds the available supply. The model is evaluated over a range of input parameters: Li supply rate, surface layer thickness, redeposition coefficient, and dissipated energy per Li particle lost to the plasma. Based on the results, first, surface temperature locking is expected above a deposited power of ∼10 MW m−2. Second, Li targets are expected to be extremely robust against power deposited during short transient events. A surface layer thickness of 50 micron is sufficient to withstand 60 MJ m−2 vertical displacement events or 20 MJ m−2 disruptions.

Comprehensive 3-D simulation and performance of ITER plasma facing and nearby components during transient events—Serious design issues

A key obstacle to a successful magnetic fusion energy production in Tokamak reactors is performance during abnormal events. Abnormal events include plasma disruptions, edge-localized modes (ELMs), vertical displacement events, and runaway electrons. While tremendous efforts are being made to find ways to mitigate such events, a credible reactor design must be able to tolerate a few of these transient events. We have recently enhanced our comprehensive HEIGHTS (High Energy Interaction with General Heterogeneous Target Systems) simulation package to enable detailed 3-D investigation of the overall aspects of plasma-material interaction processes during all the transient events. Advanced models and numerical tools were developed to efficiently couple major key processes during the transient events, and in particular disruptions and giant ELMs. These include dynamic interaction, deposition, and scattering of the escaping core plasma particles with the evolving and propagating secondary divertor vapor/plasma in the strong magnetic field. These details are critical for assessing the damage to all interior components, including the hidden structure and the first wall which were not directly exposed to these transient events and never thought to be affected as a result. Despite developing numerous efficient numerical techniques and solution methods, such calculations take several months on current supercomputers to complete. Our present results show, for the first time, that unmitigated transient events could cause significant melting and vaporization damage to most interior and hidden components, including the first wall that were not directly exposed to these events. The current ITER divertor design may not work properly and need to be significantly modified or redesigned to prevent this damage.

Simulation of electromagnetic VDE plasma effects on divertor structures of DEMO

Vertical Displacement Events (VDE’s) of plasma are one of the major concern in the design of the tokamak components inside the vacuum vessel, and in particular the divertor, because of the sideways forces generated by the interaction of the electric currents arising during VDE on the conductive metallic structures and the magnetic fields. The generation of these current is still matter of investigation either by theoretical studies and experimental campaign on the existing facilities, but it’s still lacking a generally agreed methodology to define the design inputs for the in vessel components. Further, the codes simulating plasma wall interactions make an accurate reconstruction of the plasma dynamics but suffer of a little detail in the modeling of the mechanical structures, making so the obtained results not completely reliable to define the real loads. On the other hand, the use of phenomenological correlations relating the currents to the plasma nominal parameters before the VDE lead to a wide range of possible load values, with consequent uncertainties in the design process.In this work a different approach is used in the attempt to reduce this range, modeling the mechanical structures inside the vessel with a detail and a spatial resolution suitable to verify the design, and simulating the plasma actions only as current sources via inductive coupling and direct contact, disregarding the analysis of the plasma physical quantities evolution. In this way is possible also to follow the time transient behavior related to the current diffusion in the component under study without particular assumptions on the plasma wall interactions dynamics. So in this work, after a more detailed description of the model, the results obtained for the preliminary design of the divertor cassette for DEMO are given.

Survival and in-vessel redistribution of beryllium droplets after ITER disruptions

The motion and temperature evolution of beryllium droplets produced by first wall surface melting after ITER major disruptions and vertical displacement events mitigated during the current quench are simulated by the MIGRAINe dust dynamics code. These simulations employ an updated physical model which addresses droplet-plasma interaction in ITER-relevant regimes characterized by magnetized electron collection and thin-sheath ion collection, as well as electron emission processes induced by electron and high-Z ion impacts. The disruption scenarios have been implemented from DINA simulations of the time-evolving plasma parameters, while the droplet injection points are set to the first-wall locations expected to receive the highest thermal quench heat flux according to field line tracing studies. The droplet size, speed and ejection angle are varied within the range of currently available experimental and theoretical constraints, and the final quantities of interest are obtained by weighting single-trajectory output with different size and speed distributions. Detailed estimates of droplet solidification into dust grains and their subsequent deposition in the vessel are obtained. For representative distributions of the droplet injection parameters, the results indicate that at most a few percents of the beryllium mass initially injected is converted into solid dust, while the remaining mass either vaporizes or forms liquid splashes on the wall. Simulated in-vessel spatial distributions are also provided for the surviving dust, with the aim of providing guidance for planned dust diagnostic, retrieval and clean-up systems on ITER.

Design and Analysis Progress of ITER Diagnostic Equatorial Port #09

ITER is the world's largest fusion device currently under construction in the South of France with >50 diagnostic systems to be installed inside the port plugs (PPs), the interspace (IS), or the port cell region of various diagnostic ports. The plasma facing diagnostic first wall (DFW) and its supporting diagnostic shield modules (DSM) are designed to protect front-end diagnostics from plasma neutron and plasma radiation, while providing apertures for diagnostic access to the plasma. The design of ITER port plug structures including the DFW and the DSM is largely driven by the electromagnetic loads included on the PP structural components during plasma major disruptions and the vertical displacement events (VDEs). Unlike DFW and DSM, the design of diagnostic system, however, is likely driven by the steady-state thermal loads from plasma volumetric and surface heating and the dynamic response of the in-port components attached to the port-specific DSM or closure plate under transient loads induced on the vacuum vessel and the port extension during asymmetric VDEs. Three tenant diagnostic systems are integrated into the equatorial port 09. The toroidal interferometer/polarimeter, or TIP system, is installed in the left drawer (DSM3, left looking from plasma) for measuring the plasma density so to control the fuel input. The electron cyclotron emission (ECE) system is installed in the middle drawer (DSM2) to provide the high spatial and temporal resolution measurements of electron temperature evolution and the electron thermal transport inferences. The visible/infrared wide angle viewing system is installed in the right drawer (DSM1) to provide visible and infrared viewing and temperature data of the first wall for its protection in support of machine operation. The port plug integration design and multiphysics analyses are performed following port integration requirements including the weight limit (45 tones total), shut down dose rate limits, the cooling/heating and structural integrity validation. Mass distribution for TIP and ECE DSMs has been optimized to minimize the total weight by a new design of the boron carbide shielding pocket. The lightened DSM maintains its front-end load distribution and the structural stiffness with minimum impact to the DFW so to better protect on-board diagnostics; while still provides sufficient front-end stiffness for its structural integrity as well as the diagnostics function requirements.