Divertor Area Research Articles

Optimized collection optic design for divertor Thomson scattering diagnostics in KSTAR.

Last year, the KSTAR divertor material was changed from carbon to tungsten tiles. An optimized collection optic design for divertor Thomson scattering diagnostics in KSTAR was conducted for electron temperature (1-100eV) and electron density (1 × 1018-1 × 1019m-3) profile diagnostics. This diagnostic system will utilize a 1064nm Nd:YAG laser directly from the K-top port toward the beam dump located at K-bottom, while collecting scattered light from five spatial points in the divertor area via collection optics situated in the j-middle port. Given spatial limitations, the solid angle of measurement points is limited, and the collection optic design facing the tungsten divertor is susceptible to stray light. So, the design of the collection optic is important for divertor Thomson scattering diagnostics. For optimal performance, we performed two types of collection optic designs: Cooke-triplet and double-Gaussian. We present performance ray tracing analysis results for both designs and derive the optimal design.

Performance assessment of spectroscopic measurements of hydrogen and beryllium influxes and ionization fronts in ITER divertor plasmas using the divertor impurity monitor

Characteristics of Hα and Be II emissions in ITER divertor plasmas have been evaluated, and the applicability of a spectroscopic evaluation method of influxes and ionization front positions by using the divertor impurity monitor (DIM) has been systematically verified for the first time based on plasma profile datasets with parameter scans. In addition, practical values of the ionization/photon factor (S/XB) for DIM to evaluate influxes have been determined. In attached divertor plasmas and the onset of plasma detachment, H influx integrated over the entire area of the divertor can be evaluated within ±30 % accuracy using Hα intensities measured by the DIM and S/XB value of 32, which is higher than the S/XB in most divertor plasmas in previous devices owing to high electron density. The ionization front position can also be evaluated to within a few mm by using peak channel position. However, when electron temperature further decreases from the detachment onset, the ionizing plasma component (caused by electron impact excitation) in the total Hα emission decreases and the spatial profiles of emission and ionization diverge, making spectroscopic evaluation difficult. In all conditions, Hα intensities could be measured with a more than sufficient number of photons at the required time resolution using the current DIM design. For Be influx integrated over the entire divertor area, it has been found that Be I is not suitable for the DIM since its intensity is similar to the background continuous spectra. Replacing Be I with Be II solves this problem. Even in detached divertor plasmas, it is possible to evaluate Be1+ ionization flux within ±30 % accuracy using Be II intensities and S/XB, the value of which depends on the DIM’s fields of view. However, the evaluated Be1+ ionization flux is lower than Be influx by more than 80 %, meaning that the evaluation of Be influx using Be II would be an underestimate. But, Be II can be measured with a more than sufficient number of photons.

Qualification of W heavy alloys as plasma facing material

Highly loaded divertor regions in fusion experiments are exposed to high particle and power fluxes. Therefore, tungsten with its excellent thermal and physical properties is a preferred plasma facing material for this application. For divertor areas with moderate steady-state heat flux up to 10 MW/m2 the excellent high-temperature resistance of W is not absolutely necessary. 95–97 % W heavy alloys containing Ni, Cu or Fe could be used from the aspects of plasma wall interaction and their thermomechanical properties. In this study we present results of thermo-mechanical characterisations like measurement of thermal conductivity and tensile tests at elevated temperatures as well as results of high heat flux performance tests at adiabatic and steady-state loading. Finally, we discuss results of deuterium retention measurements.

Wall conditioning effects and boron migration during boron powder injection in ASDEX Upgrade

The efficacy of boron powder injection as a wall conditioning tool method in terms of its ability to create a sufficiently uniform boron layer on plasma-facing surfaces has been studied in ASDEX Upgrade. Boron powder was injected in two series of dedicated plasma discharges at varied injection rates and total amount injected. The resulting boron deposition was determined quantitatively by exposure of witness samples followed by ex-situ surface analysis of the retrieved samples. In both experiments, isotopically enriched boron was used to distinguish the deposition of the newly injected material from the residual boron fraction in the machine originating from previous glow discharge boronisations. It could be confirmed that the injected boron is migrating and re-deposited across plasma-facing wall surfaces already within one discharge. At erosion dominated divertor areas, the boron influx from the main chamber results in formation of a mixed tungsten-boron surface layer with a boron area density of O(1nm) whereas at deposition dominated areas closed boron layers grow with ongoing boron injection to a thickness of up to O(1μm). Extrapolating the radial boron deposition profile on the samples exposed in the main chamber to the limiter front yields a similar boron coverage of O(1μm), which is about ten times higher than typical values for glow discharge boronisation. Together with the observed reduction of oxygen level and improved wall pumping, the surface analysis results demonstrate that boron powder injection provides a suitable means to refresh the wall conditioning effect of a preceding glow discharge boronisation.

Very High spatial Resolution IR thermography diagnostic positioning system upgrade in WEST Tokamak

The Very High spatial Resolution (VHR) infrared thermography diagnostic developed by CEA-IRFM is installed in the WEST tokamak to measure surface temperature of the actively cooled W-monoblocks (MB) of the Plasma Facing Units (PFUs), which are similar to the International Thermonuclear Experimental Reactor (ITER) ones. It is a major diagnostic to investigate the behavior of PFU during experimentation thanks to its very high spatial resolution (0.1 mm / pixel), movable field of view (FoV) and the wide-range temperature measurement. In particular, it can study the evolution of monoblocks mounted in the tokamak, the influence of the shaping of these components submitted to high heat load and the behavior of the leading edges regarding the assembling tolerances between adjacent monoblocks. Finally, such diagnostic directly contributes to the specification assessment of the ITER divertor units. This diagnostic was operational during the 2019 WEST experimentation and its enhanced design for the 2020 WEST experimentation campaign. In order to monitor 30 × 12 mm monoblocks temperature, the VHR IR diagnostic is able to move its narrow FoV (64 × 51 mm) to specific targets on the observable divertor area (360 × 420 mm). Position and focus steering are achieved through motor-controlled mirrors and lenses operated by a dedicated Labview application on a remote unit. Mirror positioning occurs between pulses under normal WEST operation conditions, and can withstand thermal radiation and disruptions during pulses. Repeatability, hysteresis, mechanical components reliability are critical parameters to provide accurate positioning (<1 mm) of the FoV. The tuning and calibration of these actuators in accordance with WEST environment is crucial to guarantee the best position of the FoV and optical performances. This paper gives the FoV positioning implementation details and performance of the VHR in WEST.

Double Pulse LIBS Analysis of Metallic Coatings of Fusionistic Interest: Depth Profiling and Semi-Quantitative Elemental Composition by Applying the Calibration Free Technique

In this work we report the characterization of thin metallic coatings of interest for nuclear fusion technology through the ns double-pulse LIBS technique. The coatings, composed of a tungsten (W) or tungsten-tantalum (W-Ta) mixture were enriched with deuterium (D), to simulate plasma-facing materials (PFMs) or components (PFCs) of the next generation devices contaminated with nuclear fuel in the divertor area of the vacuum vessel (VV), with special attention to ITER, whose divertor will be made of W. The double pulse LIBS technique allowed for the detection of D and Ta at low concentrations, with a single laser shot and an average ablation rate of about 110 nm. The calibration free (CF-LIBS) procedure provided a semi-quantitative estimation of the retained deuterium in the coatings, without the need of reference samples. The presented results demonstrate that LIBS is an eligible diagnostic tool to characterize PFCs with high sensitivity and accuracy, being minimally destructive on the samples, without PFCs manipulation. The CF-LIBS procedure can be used for the search for any other materials in the VV without any preliminary reference samples.

Bus Design for the Poloidal Field Coils of the NSTX-Upgrade Fusion Device

The National Spherical Torus eXperiment (NSTX) has undergone a major upgrade to NSTX-U at Princeton Plasma Physics Laboratory (PPPL). NSTX upgrade (NSTX-U) will double the toroidal field, plasma current, and neutral beam injection heating power, as well as significantly increase the pulse duration. NSTX-U uses three poloidal field (PF) coils at the vessel top and three at the bottom near the divertor areas to control the local plasma shape there. These shaping coils operate at a maximum terminal voltage of 2 kV, corresponding to a maximum current about 20 kA. To supply the power to the PF coils, hard copper bus bars are typically used near the vacuum vessel, while water cooled flexible cables are used away from the vessel. The detailed design of the hard copper bus bars will be covered in this article. During operation, these hard bus bars are subject to high electromagnetic (EM) forces, thermal displacement loads, as well as plasma halo loads due to disruption. The EM, thermal, and structural analysis were performed, and the results revealed that, with the worst loads combined, the new design meets the NSTX-U thermal, structural, and fatigue cycle requirements. The manufacturing and installation process of the bus work will be discussed at the end of this article.

First snowflake divertor experiments in MAST-U tokamak

First snowflake (SF) divertor experiments in the MAST-U tokamak demonstrated steady-state snowflake-plus divertor configurations in 450 kA ohmic L-mode plasmas. The SF divertor configuration features a second poloidal field (PF) null in the divertor region close by or overlapping with the main X-point. The resulting low PF region and two additional divertor legs (strike points) may lead to additional power and particle flux sharing via a hypothesized convective cell, and increased plasma-wetted area and radiation. The free-boundary Grad–Shafranov equilibrium code FIESTA was used to design SF configurations with several inter-null distances and orientations. In the experiment, the SF configurations with inter-null distances 0.13–0.20 m and lasting 0.2–0.3 s were obtained. Parallel connection lengths between the midplane and the outer strike point in the SF configurations (evaluated at field lines 1–2 mm from the separatrix in the midplane) were 25–30 m, higher than in the standard divertor (20–25 m) or the Super-X divertor (25 m). Diagnostic measurements highlighted salient SF features. The infra-red video bolometer diagnostic showed that the radiated power peaking in the PF null region was not as pronounces as in the standard divertor. Divertor ion fluxes measured by target Langmuir probes showed increased ion flux in the plate region where a secondary SF strike point landed, concomitantly with the SF formation. These measurements may suggest that some particle and heat redistribution was taking place in the convective SF zone. The first SF experiments provide a basis for future SF studies in MAST-U tokamak with higher input power, improved plasma control and diagnostic measurements, to be compared with the modeling predictions of plasma convective SF mixing and lower density strike point detachment threshold. • First snowflake (SF) divertor experiments in the MAST-U tokamak demonstrated steady-state snowflake-plus divertor configurations with inter-null distances 0.13–0.20 m in 450 kA ohmic L-mode plasmas. • The free-boundary Grad–Shafranov equilibrium code FIESTA was used to design SF configurations with several inter-null distances and orientations. • Divertor geometry (i.e., parallel connection lengths, divertor flux expansion and area expansion) in the SF-plus divertor fvorably comared to the standard divertor • Divertor measurements with the infra-red video bolometer diagnostic and target Langmuir probes may suggest that some particle and heat redistribution took place in the convective SF zone.

Space-resolved vacuum-ultraviolet spectroscopy for measuring impurity emission from divertor region of EAST tokamak

The measurement of impurity distribution in the divertor region of tokamaks is key to studying edge impurity transport. Therefore, a space-resolved vacuum-ultraviolet (VUV) spectrometer is designed to measure impurity emission in the divertor region on EAST. For good spectral resolution, an eagle-type VUV spectrometer with 1 m long focal length with spherical holograph grating is used in the system. For light collection, a collimating mirror is installed between the EAST plasma and the VUV spectrometer to extend the observing range to cover the upper divertor region. Two types of detectors, i.e. a back-illuminated charge-coupled device detector and a photomultiplier-tube detector, are adopted for the spectral measurement and high-frequency intensity measurement for feedback control, respectively. The angle between the entrance and exit optical axis is fixed at 15°. The detector can be moved along the exit axis to maintain a good focusing position when the wavelength is scanned by rotating the grating. The profile of impurity emissions is projected through the space-resolved slit, which is set horizontally. The spectrometer is equipped with two gratings with 2400 grooves/mm and 2160 grooves/mm, respectively. The overall aberration of the system is reduced by accurate detector positioning. As a result, the total spectral broadening can be reduced to about 0.013 nm. The simulated performance of the system is found to satisfy the requirement of measurement of impurity emissions from the divertor area of the EAST tokamak.

Development and implementation of Divertor Fast Particles Injection for EAST tokamak

The divertor is the main device in the tokamak that is subjected to heat loads, excessive high heat flux could cause damage to the divertor target. Divertor heat flux control is very important for magnetically confined fusion devices. As a method for divertor heat flux control, the Divertor Fast Particles Injection (DFPI) is successfully developed and implemented on the Experimental Advanced Superconducting Tokamak (EAST). The components of DFPI are a high-pressure gas source (0.6–1 MPa), a pulse solenoid valve, a nozzle and the pipeline etc. Compared with the Supersonic Molecular Beam Injection (SMBI) particle injection from midplane, the particles injected by DFPI are more difficult to enter the plasma core and have less impact on the performance of the core plasma; Compared with the gas puffing (GP) particle injection from the divertor, the particles injected by DFPI can enter the divertor area faster, the delay time of DFPI is about 25 ms. The successful implementation of DFPI provides an important tool for divertor heat flux control on EAST.

Predictive modelling of liquid metal divertor: from COMPASS tokamak towards Upgrade

Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict the behavior of a similar target on COMPASS Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in the future EU DEMO attached divertor. We simulate the heat conduction, sputtering, evaporation, excitation and radiation of lithium and tin in the divertor area. Measured high-resolution data from COMPASS tokamak were rescaled towards the Upgrade based on many established scalings. Our simulation then yields the amount of released metal which ranges from 4 mg s−1 upto 12 g s−1 depending mainly on the geometry and Li/Sn choice, quite independently from active cooling or strike point sweeping.

Comparing spontaneous and pellet-triggered ELMs via non-linear extended MHD simulations

Injecting frozen deuterium pellets into an ELMy H-mode plasma is a well established scheme for triggering edge localized modes (ELMs) before they naturally occur. This paper presents non-linear simulations of spontaneous type-I ELMs and pellet-triggered ELMs in ASDEX Upgrade performed with the extended MHD code JOREK. A thorough comparison of the non-linear dynamics of these events is provided. In particular, pellet-triggered ELMs are simulated by injecting deuterium pellets into different time points during the pedestal build-up described in A Cathey et al (2020 Nuclear Fusion 60 124007). Realistic ExB and diamagnetic background plasma flows as well as the time dependent bootstrap current evolution are included during the build-up to accurately capture the balance between stabilising and destabilising terms for the edge instabilities. Dependencies on the pellet size and injection times are studied. The spatio-temporal structures of the modes and the resulting divertor heat fluxes are compared in detail between spontaneous and triggered ELMs. We observe that the premature excitation of ELMs by means of pellet injection is caused by a helical perturbation described by a toroidal mode number of n = 1. In accordance with experimental observations, the pellet-triggered ELMs show reduced thermal energy losses and a narrower divertor wetted area with respect to spontaneous ELMs. The peak divertor energy fluence is seen to decrease when ELMs are triggered by pellets injected earlier during the pedestal build-up.

Completion of the production of the W7-X divertor target modules

The installation of an actively water cooled divertor in the stellarator Wendelstein 7-X (W-7X) is mandatory to achieve stationary power and particle exhaust for pulse lengths up to 30 min. The highly loaded divertor area is made of 100 target modules distributed in ten divertor units. The target modules have mechanical support frames with attachment systems to the plasma vessel, and manifolds to distribute water equally between the target elements. A target element is designed to remove a stationary heat flux up to 10 MW/m² and is made of a CuCrZr copper alloy heat sink armored with CFC NB31 tiles. The manufacturing process, assembly and quality assessment of the last 70 target modules has been successfully completed in the Integrated Technical Centre of IPP-Garching. Some parts such as the target elements and manifolds were delivered by industry. The quality was assessed as follows: visual inspections, measurement of the 3-D CFC surface, dynamic pressure tests, He leak testing under pressure at different temperature (20 °C, 160 °C) in vacuum oven, high heat flux testing. The production of the water cooled divertor is now completed, and the mounting operation of the target modules in the plasma vessel of W7-X has started.

Plasma processed tungsten for fusion reactor first-wall material

Tungsten is one of the prime candidates for a first-wall material near the divertor area due to its high temperature strength, high thermal conductivity, low erosion rate and low tritium retention. The erosion resistance of tungsten to the edge plasmas and transient events are carefully investigated in a simulated fusion environment. Here, we use the dense plasma focus (DPF) device operated in a D2 as a source for pulsed fusion plasma. The tungsten (α-W) substrates with a preferential growth direction along (110) plane were used. These pristine-W samples were nanostructurized using a (i) low-temperature continuous nitrogen RF plasma system and (ii) coated with 60 nm tungsten film, using high-temperature argon plasma in a dense plasma focus (DPF) device. The low- temperature plasma treatment created mesh-like porous nanostructure on the surface of pristine-W with change in crystalline orientation to (200), while the DPF-based deposition resulted in a nanocrystal (30–50 nm) decorated surface with enhanced (200) orientation. The crack propagation and bubble formation during DPF D2 plasma exposure were significantly controlled by the surface modification of tungsten. The mesh-like structure was modified to form loosely bound spherical nanoparticles, while the nanocrystals remained tightly bound and grew in size with D2 plasma exposure. The better adhesion of the nanocrystals and controlled growth along the (200) direction resulted in least change in hardness measurements for the nanocrystal decorated samples. Thus, nanocrystal decoration of tungsten with a preferential growth direction of (200) can help reduce the fusion-induced damage in first-wall materials.

Development of the remote handling connector for ITER divertor diagnostic system

Diagnostics is a vital system for ITER to collect information during operation. Mineral insulated cables route the electrical and optical diagnostics sensor signals from the divertor area to the diagnostic hall. The diagnostic divertor cassettes are replaced during maintenance, which raises the requirement for a connector between divertor cassette and in-vessel wall. Since it needs to be handled remotely during the installation of the 16 diagnostic divertor cassettes, it is called Remote Handling Connector. Its operating space is limited and its environmental conditions are ultra-high vacuum, baking temperature 350 °C, irradiation and challenging electromagnetic forces. The preliminary design of the Remote Handling Connector is based on the conceptual design of Outboard and Inboard Configurations and is highly impacted by finding an appropriate balance between the external and internal space limitations and system requirements. The design of the system has been an iterative process including several conceptual and mechanical design phases, thermal, magnetic and structural load analysis, risk analysis, and remote handling assessment. Mock-ups were manufactured and tested at Divertor Test Platform 2 and Remote Handling Connector Platform at VTT Tampere Finland. The developed system was evaluated at a preliminary design review meeting organized at the end of 2019. The preliminary design phase provides a baseline for the final design of the Remote Handling Connector system.

Conceptual design of CFETR divertor supporting structures based on DOF and stability theories for the remote handling requirements

The divertor area of China Fusion Engineering Test Reactor (CFETR) vacuum vessel (VV) consists of 48 modular cassettes which must be replaced during operation of the CFETR. Considering the working environment for the divertor in Tokamak, the process for replacing is conducted by remote handling (RH) ways. In this paper, the work mainly focuses on the development of a new approach to the conceptual design of CFETR Divertor RH compatible structures. The approach is based on two theories: degree of freedom (DOF) and the stability.DOF theory requires the divertor should be constraint completely, and then its DOF is equal to zero. Besides that, the stability theory requires the divertor cannot be shaken due to the projection of the centre of mass and cannot be rotated due to the larger tilting moment. Then a new design scheme for the compatible structures is put forward based on theories to fulfill the working requirements. Finally, the simulation process in Delmia is carried on in order to validate the RH feasibility.

Hydrogen permeability of AISI 316 ITER grade stainless steel

Tritium retention in the vacuum vessel emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER) several years ago. Present-day outlook reveals that the main tritium amount will be retained in deposits in the divertor area and by repeatedly applied proper thermal treatment its amount would never exceed the permitted critical mass. During such heat treatments, the vacuum vessel will be heated, too, and it may represent a tritium sink, which is highly undesirable, as the kinetics of the tritium absorption rate is not known. The presented results on permeation through AISI 316 ITER grade stainless steel membranes in the pressure range between 50 and 1000 mbar at temperatures from 200 °C to 400 °C match well with reported values on AISI 316 steels. The accent is on the influence of the residual atmosphere on achieved permeability during heat treatments at moderate temperatures.

Servo valve endurance test for water-hydraulic systems in ITER-relevant conditions

ITER Divertor maintenance equipment operates in the vacuum vessel in elevated temperature and under considerable radiation load. The heavy Divertor assemblies are lifted and transported using servo valve systems, such as Cassette Multi-functional Mover (CMM). Systems are powered with water hydraulics, using demineralized water as a pressure medium. Operations have not been tested in ITER-relevant environmental conditions and over projected duty cycles. As the hydraulic medium is rather aggressive and there are no servo valves designed for demineralized water, the greater than 2000-h operational time was considered a potential issue. Hence, a project was undertaken to ascertain the component compatibility with the environment and pressure medium, and their robustness over the required operational period. Irradiation of components was not considered at this phase of technology validation. A heated test chamber was constructed to emulate the projected maximum ambient temperature of 50 °C in the Divertor area. Test routines and measurements were specifically tailored to monitor the operational parameters of the servo valve. After the 2188-h test the servo valve parameters remained within the limits promised by the manufacturer. Pressure gain decreased and hysteresis increased but remained within the allowable limits. These changes did not have significant effect on the joint angle tracking error.

Heat management for water-hydraulic systems at ITER remote handling

The full exchange of the ITER Divertor is performed with the Divertor Remote Handling System during ITER maintenance campaigns. Equipment with high power density are required to transport the 10-tonne cassettes into and out of the vessel. Water hydraulics has proven to deliver the required power and tracking accuracy. The ambient temperature within the Divertor area (in-vessel) is between 15 °C to 55 °C, which is at the upper limit of allowed temperature range for water hydraulic components. As the Cassette Toroidal Mover (CTM) Hydraulic Power Unit (HPU) must be situated within the CTM chassis and power for the Cassette Multifunctional Mover (CMM) hydraulics is supplied from the transfer cask, they are under increased thermal load from the radioactive components. To address this, a project to investigate solutions for heat management of CTM and CMM hydraulics was undertaken. Multiple heat management methods were analysed and the best solutions for these cases were selected after trade-off analyses. As a result, a traditional liquid/air heat management solution was selected for the CMM and no dedicated cooling was selected for the CTM. Whether the solutions proposed herein are applicable to other ITER systems is discussed.

Revisiting Carbon Materials as Plasma-Facing Materials of a Fusion Reactor

Carbon materials (C) are revisited to use as a plasma-facing material (PFM) of a fusion reactor. Characteristics of C concerned, such as erosion, redeposition, H retention, and neutron damages, are reinvestigated considering recent observations in laboratories and plasma machines, in particular, JET and JT-60U. It is concluded that C can be a good PFM of a fusion reactor if they are used as armor tiles mechanically fixed to heat sink allowing possible temperature escalation of tile surfaces, which mitigate T retention and neutron irradiation effect. The mechanical fixing ensures easy replacement of the carbon armor tiles by remote handling and periodic break for reactor maintenance gives opportunity for routine replacements of the armor tiles. In addition, routine T removal by isotopic exchange with D discharges and installation of replaceable cooled plates at divertor area to take up redeposited C incorporating with T would significantly reduce T inventory in a rector.