Vertical Displacement Events Research Articles

A study on ohmic plasma initiation for JA DEMO

We have investigated ohmic plasma initiation for JA DEMO, a design concept of the tokamak demonstration fusion reactor examined in Japan. A calculation model describing the tokamak plasma initiation phase consisting of the two-dimensional analysis of the coil and eddy currents and the zero-dimensional plasma analysis is developed. It is revealed that the conducting shell does not necessarily hinder plasma initiation, while it provides the stabilizing effect of vertical displacement events. The available loop voltage and prefill gas pressure for initiation in JA DEMO are clarified for developing the initiation scenario. We demonstrate a possible scenario of ohmic plasma initiation, which is consistent with the design specifications of superconducting poloidal field coils.

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.

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.

Plasma position measurement with collimated neutron flux monitor diagnostics on JET

In future burning plasma devices measuring the plasma position and its shape with great spatial and temporal resolution is a crucial task. Nowadays these information are obtained by means of magnetic coils installed inside the vacuum vessel that in the future devices (like ITER and DEMO), due to the harsh environment caused by the high plasma temperature, will experience degradation over the time. Furthermore, the long plasma discharges will result in large integration drift causing a lack of accuracy in the measured signal. In order to assist the magnetic diagnostics and at the same time provide a novel tool to benchmark them, here the measurement of the plasma magnetic axis position by means of a collimated neutron flux monitor is proposed. Three different methods are here described and applied on JET by means of the neutron camera: a weighted average, the asymmetry method and a neural network. The methods are calibrated on a large database of plasma discharges including NBI and ICRH heated ones, and then compared with the magnetic axis position reconstructed by EFIT. The neural network outperforms the two other methods. In particular, the asymmetry method results to be sensitive to MHD activity, NBI power variation and to neutron emissivity profiles presenting a strong asymmetry, such as in case of peripheral NBI deposition due to high density plasmas or ICRH resonance layer. A possible application to vertical displacement events and disruptions is discussed and envisaged for future applications on DEMO. Finally, the performances of the neural network and of the asymmetry methods are studied in the case of one or two missing channels in the neutron flux monitor, showing how in general the reconstruction of the radial magnetic axis in both methods is more sensitive to the lack of channels than the vertical one. The methods here proposed can be applied and benchmarked on DTT and ITER neutron cameras as part of a future real-time control system.

The JOREK non-linear extended MHD code and applications to large-scale instabilities and their control in magnetically confined fusion plasmas

JOREK is a massively parallel fully implicit non-linear extended magneto-hydrodynamic (MHD) code for realistic tokamak X-point plasmas. It has become a widely used versatile simulation code for studying large-scale plasma instabilities and their control and is continuously developed in an international community with strong involvements in the European fusion research programme and ITER organization. This article gives a comprehensive overview of the physics models implemented, numerical methods applied for solving the equations and physics studies performed with the code. A dedicated section highlights some of the verification work done for the code. A hierarchy of different physics models is available including a free boundary and resistive wall extension and hybrid kinetic-fluid models. The code allows for flux-surface aligned iso-parametric finite element grids in single and double X-point plasmas which can be extended to the true physical walls and uses a robust fully implicit time stepping. Particular focus is laid on plasma edge and scrape-off layer (SOL) physics as well as disruption related phenomena. Among the key results obtained with JOREK regarding plasma edge and SOL, are deep insights into the dynamics of edge localized modes (ELMs), ELM cycles, and ELM control by resonant magnetic perturbations, pellet injection, as well as by vertical magnetic kicks. Also ELM free regimes, detachment physics, the generation and transport of impurities during an ELM, and electrostatic turbulence in the pedestal region are investigated. Regarding disruptions, the focus is on the dynamics of the thermal quench (TQ) and current quench triggered by massive gas injection and shattered pellet injection, runaway electron (RE) dynamics as well as the RE interaction with MHD modes, and vertical displacement events. Also the seeding and suppression of tearing modes (TMs), the dynamics of naturally occurring TQs triggered by locked modes, and radiative collapses are being studied.

Reassessing energy deposition for the ITER 5 MA vertical displacement event with an improved DINA model

The beryllium (Be) main chamber wall interaction during a 5 MA/1.8 T upward, unmitigated VDE scenario, first analysed in [J. Coburn et al., Phys. Scr. T171 (2020) 014076] for ITER, has been re-evaluated using the latest energy deposition analysis software. Updates to the DINA disruption model are summarized, including an improved numerical convergence for the 0D power balance, limitations on the safety factor within the plasma core, and the choice to maintain a constant plasma + halo poloidal cross-section. Such updates result in a broad halo region and higher radiated power fractions compared to previous models. The new scenario lasts for ~75 ms and deposits ~29 MJ of energy, with the radial distribution of parallel heat flux q‖r resembling an exponential falloff with an effective λq=75-198 mm. A maximum halo width wh of 0.52 m at the outboard midplane is observed. SMITER field line tracing and energy deposition simulations calculate a q⊥,max of ~83 MW/m2 on the upper first wall panels (FWP). Heat transfer calculations with the MEMOS-U code show that the FWP surface temperature reaches ~1000 K, well below the Be melt threshold. Variations of this 5 MA scenario with Be impurity densities from 0 to 3∙1019 m−3 also remain below the melt threshold despite differences in energy deposition and duration. These results are in contrast to the early study which predicted melt damage to the first wall [J. Coburn et al., Phys. Scr. T171 (2020) 014076], and emphasize the importance of accurate models for the halo width wh and the heat flux distribution q‖r within that halo width. The 2020 halo model in DINA has been compared with halo current experiments on COMPASS, JET, and Alcator C-Mod, and the preliminary results build confidence in the broad halo width predictions. Results for the 5 MA VDE are compared with those for a 15 MA equivalent, generated using the new DINA model. At the higher current, significant melting of the upper FWP is to be expected.

3D simulations of vertical displacement events in tokamaks: A benchmark of M3D-C1, NIMROD, and JOREK

In recent years, the nonlinear 3D magnetohydrodynamic codes JOREK, M3D-C1, and NIMROD developed the capability of modeling realistic 3D vertical displacement events (VDEs) including resistive walls. In this paper, a comprehensive 3D VDE benchmark is presented between these state-of-the-art codes. The simulated case is based on an experimental NSTX plasma but with a simplified rectangular wall. There are differences between the physics models and numerical methods, and the VDE evolution leads to sensitivities on the initial conditions that cannot be avoided as can be done in edge localized modes (ELM) and sawtooth simulations (due to the non-cyclical nature of VDEs). Nonetheless, the comparison serves to quantify the level of agreement in the relevant quantities used to characterize disruptions, such as the 3D wall forces and energy decay. The results bring confidence regarding the use of the mentioned codes for disruption studies, and they distinguish aspects that are specific to the models used (e.g., reduced vs full MHD models). The simulations show important 3D features for a NSTX plasma, such as the self-consistent evolution of the halo current and the origin of the wall forces. In contrast to other reduced MHD models based on an ordering in the aspect ratio, the ansatz-based JOREK reduced MHD model allows capturing many aspects of the 3D dynamics even in the spherical tokamak limit considered here.

Simulations of COMPASS vertical displacement events with a self-consistent model for halo currents including neutrals and sheath boundary conditions

The understanding of the halo current properties during disruptions is key to design and operate large scale tokamaks in view of the large thermal and electromagnetic loads that they entail. For the first time, we present a fully self-consistent model for halo current simulations including neutral particles and sheath boundary conditions. The model is used to simulate vertical displacement events (VDEs) occurring in the COMPASS tokamak. Recent COMPASS experiments have shown that the parallel halo current density at the plasma-wall interface is limited by the ion saturation current during VDE-induced disruptions. We show that usual magneto-hydrodynamic boundary conditions can lead to the violation of this physical limit and we implement this current density limitation through a boundary condition for the electrostatic potential. Sheath boundary conditions for the density, the heat flux, the parallel velocity and a realistic parameter choice (e.g. Spitzer’s resistivity and Spitzer–Härm parallel thermal conductivity) extend present VDE simulations beyond the state of the art. Experimental measurements of the current density, temperature and heat flux profiles at the COMPASS divertor are compared with the results obtained from axisymmetric simulations. Since the ion saturation current density () is shown to be essential to determine the halo current profile, parametric scans are performed to study its dependence on different quantities such as the plasma resistivity and the particle and heat diffusion coefficients. In this respect, the plasma resistivity in the halo region broadens significantly the profile, increasing the halo width at a similar total halo current.

Avoidance of vertical displacement events in DIII-D using a neural network growth rate estimator

Robust disruption avoidance techniques are critical for the development of reliable fusion reactor devices. A viable reactor will require non-disruptive, long pulse operation where simply shutting down a discharge is undesirable. To achieve such performance, the plasma must be controlled to continuously avoid hazardous regimes instead of asynchronously aborting. A recent experiment on DIII-D demonstrated for the first time real-time control of proximity to a disruptive instability boundary. In particular, the vertical growth rate, an eigenvalue that characterizes the degree of instability of the plasma's vertical position, was regulated so as not to exceed DIII-D's vertical controllability limit. The open-loop growth rate was estimated in real time on the DIII-D plasma control system using a neural network model trained with tens of thousands of DIII-D shots. The model was trained to replicate the results of RZRIG [1], a rigid displacement code for calculating the growth rate. Once trained, producing an estimate using the neural network is multiple orders of magnitude faster than RZRIG, thereby making the calculation suitable for real-time execution. The control system regulated the estimated growth rate by adjusting plasma elongation and distance to the inner wall of the vessel, and this regulation was shown to reliably avoid vertical displacement event disruptions (i.e. uncontrolled vertical oscillations) of the plasma. This work presents these experimental results, including the dynamic performance and the effectiveness of the control technique. Details are presented on the training of the neural network model, including concerns such as hyperparameter tuning and uncertainty quantification. Additionally, the methodology for embedding the neural network into the control system is discussed.

Structural assessment of the EU-DEMO WCLL Central Outboard Blanket segment under normal and off-normal operating conditions

Within the framework of the EUROfusion design activities concerning the EU-DEMO Breeding Blanket (BB) system, a research campaign has been carried out at the University of Palermo with the aim of investigating the structural behaviour of the DEMO Water-Cooled Lithium Lead (WCLL) Central Outboard Blanket (COB) segment. The assessment has been performed considering three different loading scenarios: the Normal Operation (NO), the Over-Pressurization (OP) and the Upward Vertical Displacement Event (VDE-up). In particular, NO scenario represents the loading case referring to the nominal operating conditions, whereas the OP scenario refers to the loading conditions due to an in-box LOCA accident, listed as one of the BB design basis accidental events. Lastly, the VDE-up scenario is an off-normal event reproducing the plasma disruption caused by an uncontrolled vertical motion of the plasma volume. This event brings the plasma in contact with the upper part of the plasma chamber, generating a sudden energy discharge accompanied by Electro Magnetic (EM) forces acting on the structure. The study has been carried out following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted ABAQUS v. 6.14 commercial FEM code. In particular, a detailed 3D FEM model of the whole COB segment, including the back-supporting structure and its attachment system to the vacuum vessel, has been set up. Several simulations have been run to assess the thermo-mechanical performances of the segment under the afore-mentioned loading scenarios, also taking into account the impact of the tungsten (W)-armour on the overall structural response. EM loads have been considered in all the assessed scenarios. In the first two, only magnetization forces have been taken into account, while in the VDE-up scenario Lorentz’s forces have been also taken into account. The structural response has been evaluated according to the RCC-MRx structural design rules. The obtained results are herewith presented and critically discussed.

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.

Resonant Axisymmetric Modes

Axisymmetric modes in shaped tokamak plasmas are normally associated with vertical displacement events. However, not enough attention has been given to the fact that these modes can be resonant in two different ways. Firstly, for a plasma bounded by a divertor separatrix, a generic n=0 ideal-MHD perturbation, ξ, is singular at the divertor X-point(s), where Beq · ∇ξ = 0, with Beq the equilibrium magnetic field. As a consequence, n=0 perturbations can give rise to current sheets localized along the divertor separatrix. Secondly, a feedback-stabilized n=0 mode tends to acquire an Alfvénic oscillation frequency. As a result, a resonant interaction with energetic particle orbits can lead to a new type of fast ion instability.

Preliminary design and structural assessment of DTT in-vessel coil supports

The present work regards the design methodologies for a crucial component of the new thermonuclear fusion energy machine named “Divertor Tokamak Test facility” (DTT). The purpose of this experimental facility is the investigation of the thermal exhaust issue, a critical aspect in the view of DEMO construction, and future nuclear fusion reactors for commercial use. The construction of DTT machine is starting in the ENEA Site, Frascati, Italy. The poloidal magnet system in the proposed DTT includes a central solenoid divided into 6 independent modules, plus 6 external coils. The poloidal field system also includes six copper in-vessel coils, namely two in-vessel coils for radial and vertical plasma stabilization and control, and four out of six in-vessel divertor coils for magnetic control of scrape-off layer and strike point sweeping. This work aims to provide a preliminary design of the supports for the 6 in-vessel coils considering the Electromagnetic (EM) loads due to a plasma Downward (DW) Vertical Displacement Event (VDE) at the flat-top of plasma reference single null scenario. During a DW VDE a huge amount of current flows in the in-vessel coils (circumferential direction), whose interaction with the poloidal EM field generates strong EM forces. The electromagnetic loads have been calculated with MAXFEA, a two-dimensional finite elements code for the simulation of plasma equilibrium, and then applied to homogenized in-vessel coils in the structural analysis.

ITER cold VDEs in the limit of perfectly conducting walls

Recently, it has been shown that a vertical displacement event (VDE) can occur in ITER even when the walls are perfect conductors, as a consequence of the current quench (CQ) [A. H. Boozer, Phys. Plasmas 26, 114501 (2019)]. We used the extended-MHD code M3D-C1 with an ITER-like equilibrium and induced a CQ to explore cold VDEs in the limit of perfectly conducting walls, using different wall geometries. In the case of a rectangular first wall with the side walls far away from the plasma, we obtained very good agreement with the analytical model developed by Boozer that considers a top/bottom flat-plates wall. We show that the solution in which the plasma remains at the initial equilibrium position is improved when bringing the side walls closer to the plasma. When approximating the ITER first wall as a perfect conductor, the plasma remains stable at the initial equilibrium position far beyond the value predicted by the flat-plates wall limit. When considering an opposite limit in which only the inner shell of the ITER vacuum vessel acts as a perfect conductor, the plasma is displaced during the CQ, but the edge safety factor remains above 2 longer in the current decay compared to the flat-plates wall limit. In all the simulated cases, the VDE is found to be strongly dependent on the plasma current, in agreement with a similar finding in the flat-plates wall limit, showing an important difference with hot VDEs in which the CQ is not a necessary condition.

A mathematical model for calculation of the influence of ferromagnetic components in Vertical Displacement Events and stability simulations of tokamak plasmas

Iron core transformer tokamaks have a distinct element, namely the iron core which can affect the plasma stability of the plasma. As a consequence of the high non-linear dependence of the magneto-hydrodynamic solutions on the iron permeability μ Fe in JET tokamak for example, Vertical Displacement Events (VDEs), equilibrium and stability calculations are more complicated and more time consuming than in air core transformer tokamaks. By considering the ferromagnetic components as a linear, isotropic and homogeneous media on subdomains, it is known that these media can be replaced by a homogeneous one (vacuum) and a surface-current density distribution i Fe(l) on the separation surfaces between subdomains, with l the curve taken along the curve separating two different magnetic media. For the case of geometry with rotational symmetry, this surface-current density distribution along a curve, in a meridian plane, is given by a Fredholm integral equation of second kind. Practically, the advantage of this method is more obvious for the inverse formulation of the VDEs and stability problems by moving the non-linear term (due to the iron permeability μ Fe ) from the differential operator to the r.h.s. of the equations. In this paper, we are presenting how the influence of ferromagnetic components in the equations of the surface currents developed in the vessel structures during Wall Touching Kink Modes (WTKMs) can be taken into account and are reviewing the equations to be solved in order to simulate the influence of the ferromagnetic components in VDEs and equilibrium stability calculations. The numerical results of these simulations for a real JET tokamak structure and plasma parameters will be reported in a future paper.

The influence of boundary and edge-plasma modeling in computations of axisymmetric vertical displacement

A number of previously published studies apply extended-magnetohydrodynamics (MHD) computations to model macroscopic dynamics of tokamak vertical displacement events (VDEs). The characteristic plasma-wall contact and resistive-wall diffusion imply sensitivity to boundary conditions in a general sense, but which conditions have significant influence depends on how the plasma is modeled. This work examines these dependencies by comparing results of axisymmetric extended-MHD computations with different sets of boundary conditions, plasma thermal-conduction models, and wall resistivity values. The geometry and plasma parameters of the computations represent a modest-sized tokamak. The forced-displacement scenario and computational setup are similar to those used in previous 3D computations [C. R. Sovinec and K. J. Bunkers, Phys. Plasma Controlled Fusion 61, 024003 (2019)]. The results show that for a given wall resistivity, the VDE time scale is most sensitive to variations in the boundary and thermal-conduction parameters that affect electron thermal transport. The electrical conductivity depends on electron temperature, and the dependence on thermal transport stems from its influence on the electrical circuit that includes the open-field halo current. Conditions that lead to hotter, broader halo regions slow the evolution. Significant sensitivity to the boundary condition on plasma flow-velocity exists when electron thermal conduction is restricted and electron energy loss is convective, which is expected for conditions at the entrance of the magnetic presheath.

Simulation of EM loads acting on the plasma-facing units of ITER divertor cassette

Some aspects of electromagnetic (EM) simulation of the central divertor cassettes (CDC) in the ITER tokamak are described. The study is focused on EM loads anticipated on the plasma-facing units (PFU) of CDC. The paper is devoted to the EM simulations of the most dangerous scenarios associated with high mechanical stresses: fast downward vertical displacement events of Categories II and III with 36-ms linear current quench (FD VDE-II and FD VDE-III with 36-ms LCQ). The selected events have been analyzed using the original TORNADO code intended for simulation of transient EM processes in 3D solids using the finite element representation. Results of computations include evolutions of EM loads on PFUs in the form of integral EM forces and moments, their peak values and relevant time points.

On the axisymmetric stability of tokamaks with ferromagnetic walls

Reduced activation ferritic steels are an attractive option for use in large structural components surrounding tokamak plasmas in future fusion power plants, but their ferromagnetic response to the confining magnetic fields must be properly understood. Simultaneously, the advantages of operating at high plasma elongation push tokamak designs toward scenarios that are more vulnerable to vertical displacement events. Passive conducting structures in present tokamaks slow these instabilities such that they may be feedback controlled, but the efficacy of this process is likely to be eroded by ferromagnetic effects. We approach two related analytical models—in cylindrical and spherical geometries—which qualitatively and quantitatively assess the impact of a ferritic steel wall on the vertical instability growth rate for a plasma of certain elongation. Distinct limits for magnetically thick and thin walls give key physical insight, but the dependence on magnetic permeability and wall geometry is, in general, quite complex. Equilibrium considerations, particularly with respect to radial force balance, are also encountered.

Electromagnetic and mechanical analyses of the ITER electron cyclotron Upper launcher steering M4 mirrors for the vertical displacement event

Abstract The mm-wave power exiting the eight In-Vessel waveguides inside the ITER Electron Cyclotron Upper Launcher is reflected by three fixed mirror sets and finally aimed by two independent steering mirrors to specific plasma locations. A design solution for both Upper and Lower Steering M4 Mirrors capable to withstand the loads taking place during normal operation was already proposed. This design is now assessed against the Vertical Displacement Event Category III (VDEIII) linear decay (36 ms) scenario, which was identified as one of the most stringent load combinations for the steering M4 mirrors. This paper reports the analyses performed to quantify the mechanical loads induced in the steering M4 mirrors during the VDEIII scenario as well as the structural integrity assessment to validate the design against these loads. The transient electromagnetic simulations show that the highest forces are induced in the reflecting surfaces, which produce a bending moment that tends to rotate upwards the mirrors. The comparison between the categorized stresses obtained from the transient mechanical analyses and the allowable design limits according to the ASME code shows that the design of the steering M4 mirrors is widely capable of withstanding the expected loads taking place during the VDEIII scenario.

Characterization of disruption halo current between ‘W-Like’ graphite divertor and ‘ITER-Like’ divertor structure on EAST tokamak

The general characteristics of disruption halo currents, including current flowing paths between the components, are studied in the experiments of EAST tokamak with ‘W-Like’ graphite divertor and ‘ITER-like’ tungsten divertor. It is found that lower hybrid waves can efficiently reheat minor disruption plasma, this can drive plasma current for tens, or even hundreds, of milliseconds. The reheated plasma can lead to lower loop voltage and of course lower halo current. Additionally, it is confirmed that helium can not rapidly cool down plasmas with high thermal energy, but Argon can shut down plasma rapidly even with a little amount of gas. Until now no runaway electron plateau has been observed both in Argon and Helium mitigation experiments. The EAST disruption database shows that: (1) Larger loop voltage is driven in vertical displacement event disruptions than in major disruptions. (2) Halo currents decrease with the decrease of vertical displacement. (3) The boundary of is 0.72, less than the limit shown in IDDB. The experimental results described here can help to improve the design of the planned upgrade of the EAST lower divertor, and provide physics information for the ITER divertor.