Inboard Limiter Research Articles

Off-target gradient-driven flows in 3D simulations of ADITYA-Upgrade tokamak scrape-off layer plasma transport

Coupled plasma-neutral transport simulations are performed on ADITYA-Upgrade tokamak scrape-off layer (SOL) plasma, where flows in the core and SOL were measured to reverse signs with density variation. The simulations performed using the EMC3-Eirene plasma-neutral code combination incorporate the toroidally continuous high-field-side belt limiter placed in a moderate circular tokamak equilibrium. The development of mutually counter-propagating toroidal plasma flows in the top and bottom regions of both the SOL and core is recovered for relatively high upstream density cases with high input power (300 kW and 3 m2 s−1). The origin of the flows is traced to the poloidal density variation introduced by high recycling on the inboard localized belt limiter. The results are compared with similar observations, for example, in Doppler-shifted passive charge exchange line emission on the ADITYA-Upgrade (ADITYA-U) tokamak, highlighting the role played by residual stress in the total Reynolds stress. The external stimuli, such as a localized gas puff, are discussed as potential drivers of flow, via residual stress, based on the existing resonant model of the tokamak plasma rotation.

Conceptual design of poloidal horseshoe limiter layout for JA DEMO

In the development of the JA DEMO fusion reactor, the proposed poloidal horseshoe limiter aims to mitigate the surface heat load localized on the breeding blanket. The limiter protrudes from the first wall toward the plasma to shade the blanket from the charged particles, which account for the majority of the surface heat load. To concretize the limiter design, this study presents a limiter layout that implements a concept that allows for sufficient cooling and remote maintenance. The limiter shape was determined by evaluating the surface heat load using 3-D magnetic field line tracing. A fully poloidal limiter with a width of 300 mm, which corresponds to 2.4 % of the FW area, was selected. The limiter heating, consisting of surface heat load and nuclear heating, was evaluated to discuss the required cooling, and the concept of a limiter cassette and modules for cooling was proposed. For the limiter replacement method, the outboard limiter is designed to be extracted from the upper port with the mounting segment. The inboard limiter is then grasped and transported to reach the upper port after the outboard limiter extraction. Both the inboard and outboard limiter are expected to be replaced without affecting other in-vessel components in this study.

Sustainable internal transport barrier discharge at KSTAR

We explore the inboard-limited internal transport barrier (ITB) as an alternative advanced operation scenario for KSTAR. This paper presents in detail the progress of the ITB experiment at KSTAR. In an earlier study, the ITB formed in both ion and electron thermal channels, and an early neutral-beam injection (NBI) power of over 4 to 5 MW under a limited L-mode was crucial to ITB access. In the present study, we access the ITB experimentally with about 3 MW of NBI power by using the upper single null (USN), which is an unfavorable H-mode condition with a higher L–H power threshold. Finding an ITB access condition with a lower heating power should allow for a more efficient control of the heat flux and for maintaining stable plasma performance. The key control parameters of the experiment are the vertical position and the location of outboard striking point of the plasma. The shape-control attempts to divert the plasma to a vertically shifted USN with a marginal touch of the inboard limiter so that the plasma can remain in the L-mode at the boundary, while the striking-point control maintains the ITB performance for a longer period of time.

Control of electron acceleration process during merging start-up of spherical tokamak

The axial plasma merging method is one of the start-up schemes to form spherical tokamak (ST) plasma without use of a center-solenoid coil. It involves a magnetic reconnection process in the presence of high toroidal magnetic field parallel to the reconnection electric field and has the potential to accelerate electrons directly along the magnetic field lines. In the paper, the role of boundary conditions of the reconnection downstream region was investigated in the UTST experiment. Self-generated axial electric field in the reconnection downstream region was observed to grow up to 20–30 times as large as the toroidal electric field where the MHD condition holds, but was largely suppressed inside the area where the magnetic field lines are contacted to the inboard limiter conductors equipped on the center stack. This suppression effect was attributed to the short circuit connection of the magnetic field lines to hinder the axial charge separation and then a significant part of the parallel electric field remained. These experimental results suggest that the merging start-up method has the capability to enhance electron acceleration efficiency by magnetic reconnection and may have the potential to produce high electron temperature ST plasma applicable for additional heating and current drive techniques.

Features of near and far scrape-off layer heat fluxes on the Wendelstein 7-X inboard limiters

The power fall-off length is used as a characteristic dimension to describe the power exhaust in the scrape-off layer (SOL) in magnetic confinement fusion devices. Measurements from the limiter campaign of the stellarator Wendelstein 7-X with inboard limiters are presented with the first-time characterization of the limiter heat loads. Two fall-off regimes are found with a narrow near SOL with power fall-off length in millimeters and a wider far SOL with fall-off lengths of several centimeters. An attempt is made to describe both regimes with a scaling law for plasma with different heating powers and densities. The results confirm the major geometry effects of the connection length on the heat transport predicted by 3D modeling.

First divertor physics studies in Wendelstein 7-X

The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment, which went into operation in 2015, has been operating since 2017 with an un-cooled modular graphite divertor. This allowed first divertor physics studies to be performed at pulse energies up to 80 MJ, as opposed to 4 MJ in the first operation phase, where five inboard limiters were installed instead of a divertor. This, and a number of other upgrades to the device capabilities, allowed extension into regimes of higher plasma density, heating power, and performance overall, e.g. setting a new stellarator world record triple product. The paper focuses on the first physics studies of how the island divertor works. The plasma heat loads arrive to a very high degree on the divertor plates, with only minor heat loads seen on other components, in particular baffle structures built in to aid neutral compression. The strike line shapes and locations change significantly from one magnetic configuration to another, in very much the same way that codes had predicted they would. Strike-line widths are as large as 10 cm, and the wetted areas also large, up to about 1.5 m2, which bodes well for future operation phases. Peak local heat loads onto the divertor were in general benign and project below the 10 MW m−2 limit of the future water-cooled divertor when operated with 10 MW of heating power, with the exception of low-density attached operation in the high-iota configuration. The most notable result was the complete (in all 10 divertor units) heat-flux detachment obtained at high-density operation in hydrogen.

Towards a new image processing system at Wendelstein 7-X: From spatial calibration to characterization of thermal events.

Wendelstein 7-X (W7-X) is the most advanced fusion experiment in the stellarator line and is aimed at proving that the stellarator concept is suitable for a fusion reactor. One of the most important issues for fusion reactors is the monitoring of plasma facing components when exposed to very high heat loads, through the use of visible and infrared (IR) cameras. In this paper, a new image processing system for the analysis of the strike lines on the inboard limiters from the first W7-X experimental campaign is presented. This system builds a model of the IR cameras through the use of spatial calibration techniques, helping to characterize the strike lines by using the information given by real spatial coordinates of each pixel. The characterization of the strike lines is made in terms of position, size, and shape, after projecting the camera image in a 2D grid which tries to preserve the curvilinear surface distances between points. The description of the strike-line shape is made by means of the Fourier Descriptors.

Observation of Marfes in the Wendelstein 7-X stellarator with inboard limiters

In the first operational campaign of the Wendelstein 7-X (W7-X) stellarator we observed Marfe-like radiation belts at the plasma edge. We describe and analyze the behaviour of these radiation belts, especially with respect to the high density inside the belt, with a set of suitable diagnostics. We present evidence suggesting that these Marfe-like radiation belts are triggered by a drop of the edge temperature below 30 eV. From this observation we draw the conclusion that the physical cause of Marfes in tokamaks and Marfe-like radiation belts in W7-X is the same, namely radiative condensation. Furthermore, both exhibit the device symmetry, i.e. axisymmetry in a tokamak and the fivefold stellarator symmetry in W7-X.

Pre-ionization by AC Ohmic coil operation in the TST-2 spherical tokamak

AC Ohmic coil operation experiments with frequencies up to 10 kHz were performed on the TST-2 spherical tokamak device, and the pre-ionization process was studied in detail. The minimum loop voltage for pre-ionization was 0.4 V, which corresponds to 0.5 V at the inboard limiter. Dependences of growth rate and saturation level of the process on various parameters were obtained, and they are compared with a time-dependent 0-dimensional model based on Townsend avalanche and loss along field lines. Most of the dependences are reproduced qualitatively by the model, and quantitative differences are within a factor of several. However, the external vertical field dependence of the appearance time, which is defined as the time to observe a plasma, and the isotope effect cannot be reproduced by the model. An ambipolar diffusion state which is predicted theoretically but mitigated experimentally is discussed. It was found that secondary electron emission at the limiter surfaces is a candidate mechanism to mitigate the state.

Reconstruction of recycling flux from synthetic camera images, evaluated for the Wendelstein 7-X startup limiter

The interpretation of spectroscopic measurements in the edge region of high-temperature plasmas can be guided by modeling with the EMC3-EIRENE code. A versatile synthetic diagnostic module, initially developed for the generation of synthetic camera images, has been extended for the evaluation of the inverse problem in which the observable photon flux is related back to the originating particle flux (recycling). An application of this synthetic diagnostic to the startup phase (inboard) limiter in Wendelstein 7-X (W7-X) is presented, and reconstruction of recycling from synthetic observation of emission is evaluated. It is shown that recycling—in terms of re-ionization within the view cone of the camera—is underestimated by a factor of about 3 at the peak location and overestimated in between, and that emission related to molecular dissociation can locally yield an additional factor of 2 for the ionization to photon conversion. Fitted conversion factors are extracted from the simulation results which include a dependence on the plasma density.

3D Monte-Carlo study of toroidally discontinuous limiter SOL configurations of Aditya tokamak

The plasma-neutral transport in the scrape-off layer (SOL) region formed by toroidally discontinuous limiters deviates from usual uniform SOL approximations when 3D effects caused by limiter discreteness begin to dominate. In an upgrade version of the Aditya tokamak, originally having a toroidally localized poloidal ring-like limiter, the newer outboard block and inboard belt limiters are expected to have smaller connection lengths and a multiple fold toroidal periodicity. The characteristics of plasma discharges may accordingly vary from the original observations of large diffusivity, and a net improvement and the stability of the discharges are desired. The estimations related to 3D effects in the ring limiter plasma transport are also expected to be modified and are updated by predictive simulations of transport in the new block limiter configuration. A comparison between the ring limiter results and those from new simulations with block limiter SOL shows that for the grids produced using same core plasma equilibrium, the modified SOL plasma flows and flux components have enhanced poloidal periodicity in the block limiter case. These SOL modifications result in a reduced net recycling for the equivalent edge density values. Predictions are also made about the relative level of the diffusive transport and its impact on the factors limiting the operational regime.

Confinement in Wendelstein 7-X limiter plasmas

Observations on confinement in the first experimental campaign on the optimized Stellarator Wendelstein 7-X are summarized. In this phase W7-X was equipped with five inboard limiters only and thus the discharge length restricted to avoid local overheating. Stationary plasmas are limited to low densities <2–3 · 1019 m−3. With the available 4.3 MW ECR Heating core T e ~ 8 keV, T i ~ 1–2 keV are achieved routinely resulting in energy confinement time τ E between 80 ms to 150 ms. For these conditions the plasmas show characteristics of core electron root confinement with peaked T e-profiles and positive E r up to about half of the minor radius. Profiles and plasma currents respond to on- and off-axis heating and co- and counter ECCD respectively.

Measurement of inner wall limiter SOL widths in KSTAR tokamak

Scrape-off layer (SOL) widths λq are presented from the KSTAR tokamak using fast reciprocating Langmuir probe assembly (FRLPA) measurements at the outboard mid-plane (OMP) and the infra-Red (IR) thermography at inboard limiter tiles in moderately elongated (κ = 1.45 – 1.55) L-mode inner wall-limited (IWL) plasmas under experimental conditions such as BT = 2.0 T, PNBI = 1.4 – 1.5 MW, line averaged densities 2.5 – 5.1 × 1019 m−3 and plasma current Ip = 0.4 − 0.7 MA. There is clear evidence for a double exponential structure in q||(r) from the FRLPA such that, for example at Ip = 0.6 MA, a narrow feature, λq,near (=3.5 mm) is found close to the LFCS, followed by a broader width, λq,main (=57.0 mm). Double exponential profiles (λq,near = 1.5 – 2.8 mm, λq,main = 17.0 – 35.0 mm) can be also observed in the IR heat flux mapped to the OMP throughout the range of Ip investigated. In addition, analysis of SOL turbulence statistics obtained with the FRLPA shows high relative fluctuation levels and positively skewed distributions in electron temperature and ion particle flux across the SOL, with both properties increasing for longer distance from the LCFS, as often previously observed in the tokamaks. Interestingly, the fluctuation character expressed in terms of spectral distributions remains unchanged in passing from the narrow to the broad SOL heat flux channel.

Performance and properties of the first plasmas of Wendelstein 7-X

The optimized, superconducting stellarator Wendelstein 7-X went into operation and delivered first measurement data after 15 years of construction and one year commissioning. Errors in the magnet assembly were confirmend to be small. Plasma operation was started with 5 MW electron cyclotron resonance heating (ECRH) power and five inboard limiters. Core plasma values of keV, keV at line-integrated densities were achieved, exceeding the original expectations by about a factor of two. Indications for a core-electron-root were found. The energy confinement times are in line with the international stellarator scaling, despite unfavourable wall conditions, i.e. large areas of metal surfaces and particle sources from the limiter close to the plasma volume. Well controlled shorter hydrogen discharges at higher power (4 MW ECRH power for 1 s) and longer discharges at lower power (0.7 MW ECRH power for 6 s) could be routinely established after proper wall conditioning. The fairly large set of diagnostic systems running in the end of the 10 weeks operation campaign provided first insights into expected and unexpected physics of optimized stellarators.

A high resolution IR/visible imaging system for the W7-X limiter.

A high-resolution imaging system, consisting of megapixel mid-IR and visible cameras along the same line of sight, has been prepared for the new W7-X stellarator and was operated during Operational Period 1.1 to view one of the five inboard graphite limiters. The radial line of sight, through a large diameter (184 mm clear aperture) uncoated sapphire window, couples a direct viewing 1344 × 784 pixel FLIR SC8303HD camera. A germanium beam-splitter sends visible light to a 1024 × 1024 pixel Allied Vision Technologies Prosilica GX1050 color camera. Both achieve sub-millimeter resolution on the 161 mm wide, inertially cooled, segmented graphite tiles. The IR and visible cameras are controlled via optical fibers over full Camera Link and dual GigE Ethernet (2 Gbit/s data rates) interfaces, respectively. While they are mounted outside the cryostat at a distance of 3.2 m from the limiter, they are close to a large magnetic trim coil and require soft iron shielding. We have taken IR data at 125 Hz to 1.25 kHz frame rates and seen that surface temperature increases in excess of 350 °C, especially on leading edges or defect hot spots. The IR camera sees heat-load stripe patterns on the limiter and has been used to infer limiter power fluxes (∼1-4.5 MW/m2), during the ECRH heating phase. IR images have also been used calorimetrically between shots to measure equilibrated bulk tile temperature, and hence tile energy inputs (in the range of 30 kJ/tile with 0.6 MW, 6 s heating pulses). Small UFO's can be seen and tracked by the FLIR camera in some discharges. The calibrated visible color camera (100 Hz frame rate) has also been equipped with narrow band C-III and H-alpha filters, to compare with other diagnostics, and is used for absolute particle flux determination from the limiter surface. Sometimes, but not always, hot-spots in the IR are also seen to be bright in C-III light.

The Set of Diagnostics for the First Operation Campaign of the Wendelstein 7-X Stellarator

Wendelstein 7-X (W7-X) is a large optimized stellarator (B=2.5T, V=30m3) aiming at demonstrating the reactor relevance of the optimized stellarators. In 2015 W7-X will begin its first operation phase (OP1.1) with five inertially cooled inboard limiters made of graphite. Assuming the heat loads can be spread out evenly between the limiters, 1 second discharges at 2 MW of ECRH heating power could be run in OP1.1. The expected plasma parameters will be sufficient to demonstrate the readiness of the installed diagnostics and even to run a first physics program. The diagnostics available for this first operation phase, including some special limiter diagnostics, and their capabilities are being presented.A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

Self organization of high βp plasma equilibrium with an inboard poloidal magnetic field null in QUEST

Successful production of high βp plasmas (εβp ⩾ 1) fully non-inductively (NI) and their long pulse sustainment with the help of modest power (<100 kW) of electron cyclotron waves is demonstrated. High βp plasmas are found for the first time to be naturally self organized to form a stable natural inboard poloidal field null (IPN) equilibrium. A critical βp value is identified, which defines the transition boundary from inboard limiter (IL) to IPN equilibrium. A new feature of plasma self organization is evidenced, which enhances its negative triangular shape to sustain high βp. These results show a relatively simple method to produce and sustain high βp plasma close to the equilibrium limit in a stable configuration exploiting its self organization property.

The comparison of heat flux pattern on lower divertor in KSTAR

The heat flux in KSTAR is estimated for various discharge conditions by using thermocouple arrays. The heat flux on the divertor is higher than that on inboard limiter or passive stabilizer by a factor of 2. Although the plasma configuration in KSTAR has been set to a double-null configuration, the heat flux on lower divertor is higher than that on upper divertor by 3–8 times, indicating a lower-single-null-like configuration. It is observed that the operation of the in-vessel cryo-pump (IVCP) changes the heat flux pattern significantly: When the IVCP was not operated, the heat fluxes on inboard divertor (ID), central divertor (CD) and outboard divertor (OD) were similar, but when the IVCP was operated, the heat fluxes on ID and CD were increased slightly and that on OD was decreased by 2–3 times. The heat flux on divertor was decreased from 35 to 26kW/m2 with the use of the resonant magnetic perturbation (RMP), especially that on CD was decreased by 2–4 times, while that on OD is increased by 2–3 times than without RMP. For the longest H-mode pulse of 22s shot, the heat flux on lower OD was 73kW/m2, which is the maximum heat flux among the shots obtained in 2013 campaign.

Impact of a narrow limiter SOL heat flux channel on the ITER first wall panel shaping

The inboard limiters for ITER were initially designed on the assumption that the parallel heat flux density in the scrape-off layer (SOL) could be approximated by a single exponential with decay length λq. This assumption was found not to be adequate in 2012, when infra-red (IR) thermography measurements on the inner column during JET limiter discharges clearly revealed the presence of a narrow heat flux channel adjacent to the last closed flux surface. This near-SOL decay occurs with λq ∼ few mm, much shorter than the main SOL λq, and can raise the heat flux at the limiter apex a factor up to ∼4 above the value expected from a single, broader exponential. The original logarithmically shaped ITER inner wall first wall panels (FWPs) would be unsuited to handling the power loads produced by such a narrow feature. A multi-machine study involving the C-Mod, COMPASS, DIII-D and TCV tokamaks, employing inner wall IR measurements and/or inner wall reciprocating probes, was initiated to investigate the narrow limiter SOL heat flux channel. This paper describes the new results which have provided an experimental database for the narrow feature and presents an ITER inner wall FWP toroidal shape optimized for a double-exponential profile with λq = 4 (narrow feature) and 50 mm (main-SOL), the latter also derived from a separate multi-machine database constituted recently within the International Tokamak Physics Activity. It is shown that the new shape allows the power handling capability of the original shape design to be completely recovered for a wide variety of limiter start-up equilibria in the presence of a narrow feature, even taking assembly tolerances into account. It is, moreover, further shown that the new shape has the interesting property of both mitigating the impact of the narrow feature and resulting in only a very modest increase in heat load, compared to the current design, if the narrow feature is not eventually found on ITER.

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630kA, plasma duration up to 12s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3s pulse durations and Ip of 611kA) is 0.01MW/m2 while that in 2011 campaign (with 12s pulse duration and Ip of 630kA) is about 0.02MW/m2. The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.