Tungsten Tiles Research Articles

LIA-QMS method for the quantity analysis of the hydrogen isotopes retention in first-wall components of Globus-M2 tokamak

The technique based on quadrupole mass spectrometry of the material ejected by laser-induced ablation – LIA-QMS is proposed as a tool to measure H/D/T content in carbon co-deposits in the first wall and divertor regions of tokamak reactors. Tungsten tiles exposed in Globus-M2 tokamak were taken for validation experiments. Measurement accuracy was determined by comparison with the results obtained by laser-induced desorption – LID-QMS and conventional thermal desorption spectroscopy (TDS). Total amount of deuterium in hydrocarbon deposits was found to be 2.8 × 1017 D/cm2, when measured by – LIA-QMS, 3.0 × 1017 D/cm2 for LID-QMS and 2.9 × 1017 D/cm2 for TDS. The main uncertainty of D surface concentration are caused by inhomogeneity of the deposit thickness over the sample area or by presence of hydrocarbon species in the released gas.

Laser speckle interferometry approach for measuring the first wall surface micro-morphology in fusion devices with only single interferogram based on Hilbert transform

During the operation of the EAST device, the surface morphology of the first wall in will change due to the erosion and redeposition from plasma-wall interaction (PWI). In this paper, a novel phase extraction approach using only single interferogram based on twice Hilbert transform (THT) has been proposed and implemented. The current approach has a significant advance for eliminating the phase-shifting errors in principle, but the problems of phase flipping and loss due to ignoring non-physical negative frequency spectrum must to be overcome. To address these problems a judgement method is proposed in current investigation to distinguish the regions of the negative spatial frequency. Then a binary matrix was developed to modify the wrapped phase of the measured interferogram. The validation of the proposed THT approach has been demonstrated by the micro-morphology reconstruction of laser ablation craters on molybdenum (Mo) and tungsten (W) tiles, which are commonly used as the first wall in EAST device. Results indicate that the THT approach has a reasonable measurement accuracy with relative errors around 10%. Also, the current THT approach presents an enhanced robust under the EAST-like vibrational environments.

Helium retention feature in the boron deposited layer on tungsten substrate by laser-induced breakdown spectroscopy and machine learning approach

ITER is designed for a burning plasma operation in which Tungsten (W) tiles are used as the first wall and diverter materials. Studying He dynamics in B-coated W wall is essential to understanding the effect of boronized plasma-facing components in fusion reactors, as these post-conditioning materials significantly influence helium retention and release. Plasma-wall interactions (PWI) are an important issue in ITER fusion reactors. PWI would lead to wall erosion and impurity redeposition. As a product of the D-T burning plasma, helium (He) ash would be retained or co-deposited on plasma-facing components (PFCs), affecting the stable operation of burning plasma. Laser-induced breakdown spectroscopy (LIBS) is proposed as a promising in-situ diagnostic approach for monitoring D/T and He retention and impurity deposition on PFCs of tokamak devices. In this study, the LIBS technique was used to investigate the helium retention feature in the boron (B) layer on tungsten substrate in 10−5 mbar. Five He-retention samples on the boron deposition layer on tungsten substrates were prepared by pulsed laser deposition (PLD) method at different ambient pressures in our lab. The investigations indicate that the atomic spectral line of helium (He-I 587.56 nm) was observed in the spectra of the first three laser shots. The depth profiles of He, B, and W in the boron-deposited layer on tungsten substrate were performed by LIBS to determine the co-deposition layer thickness. The concentration of He in the co-deposition layer samples measured by TDS is 7 × 1020 He/m2. The plasma parameters, such as plasma electron temperature and electron number density, were calculated to validate the local thermodynamic equilibrium. Machine learning (ML) algorithms are used to classify the co-deposits and substrates. The first three principal components (PC1, PC2 & PC3) of the unsupervised ML algorithm (PCA) give a classification accuracy of 91.7 %. The supervised ML algorithm neural network achieved training and testing accuracy of 100 % and 96.7 %, respectively.

Overview of damage to beryllium limiters by unmitigated disruptions and runaway electrons in the JET tokamak with metal walls

Abstract The Joint European Torus (JET) fusion reactor was upgraded to the metallic wall configuration in 2011 which consisted of bulk beryllium (Be) tiles in the main chamber and bulk tungsten (W) and W-coated CFC tiles in the divertor [1]. During each campaign, a series of wall damages were observed; on the upper dump plates (UDP) positioned to the top part of the vessel walls and on the inner wall - mainly affecting the inner wall guard limiters (IWGL). In both cases, it was concluded that the causes of these damages were unmitigated plasma disruptions. In the case of JET with the metallic wall configuration, most of these plasma disruptions were intentionally provoked. The overall objective was to study the behaviour of these phenomena, in order to assess their impact on the wall, improve understanding of morphological material changes, and – based on that - to develop, implement and test mitigation techniques for their prospective use on ITER. The current results bring additional information on the effects of the unmitigated plasma disruptions on the UDPs and are a significant extension of work presented in [2] where the scale of the damage after three operational campaigns on the Be top limiters of JET was highlighted. In addition, new data is presented on the damaging effect that the high energetic runaway electrons had on the Be IWGL in JET.

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.

Plasmoid drift and first wall heat deposition during ITER H-mode dual-SPIs in JOREK simulations

The heat flux mitigation during the thermal quench (TQ) by the shattered pellet injection (SPI) is one of the major elements of disruption mitigation strategy for ITER. It’s efficiency greatly depends on the SPI and the target plasma parameters, and is ultimately characterised by the heat deposition on to the plasma facing components. To investigate such heat deposition, JOREK simulations of neon-mixed dual-SPIs into ITER baseline H-mode and a ‘degraded H-mode’ with and without good injector synchronization are performed with focus on the first wall heat flux and its energy impact. It is found that low neon fraction SPIs into the baseline H-mode plasmas exhibit strong major radial plasmoid drift as the fragments arrive at the pedestal, accompanied by edge stochasticity. Significant density expulsion and outgoing heat flux occurs as a result, reducing the mitigation efficiency. Such drift motion could be mitigated by injecting higher neon fraction pellets, or by considering the pre-disruption confinement degradation, thus improving the radiation fraction. The radiation heat flux is found to peak in the vicinity of the fragment injection location in the early injection phase, while it relaxes later on due to parallel impurity transport. The overall radiation asymmetry could be significantly mitigated by good synchronization. Time integration of the local heat flux is carried out to provide its energy impact for wall heat damage assessment. For the baseline H-mode case with full pellet injection, melting of the stainless steel armour of the diagnostic port could occur near the injection port, which is acceptable, without any melting of the first wall tungsten tiles. For the degraded H-mode cases with quarter-pellet SPIs, which have 1/4 total volume of a full pellet, the maximum energy impact approaches the tolerable limit of the stainless steel with un-synchronized SPIs, and stays well below such limit for the perfectly synchronized ones.

Exploring of two connection forms of the tungsten tile and CuCrZr heat sink for first wall mockup

The first wall (FW) of a blanket or divertor faces high temperature plasma. Thus, the FW needs to withstand extremely high heat fluxes. Periodic plasma pulses bring some reliability challenges to FW structures. The FW is composed of tiles, heat sink and a bolt with different connection between the tile and the heat sink. In this study, the performances of two connection forms for the tile and the heat sink were investigated. The FLUENT software was used to evaluate the heat transfer performances for the two types structure. Elastic-plastic analysis was subsequently adopted to estimate the structural properties based on the thermal hydraulic results. At the same time, the thermal and structural results under a transient high thermal load (2 MW/m2) were also analyzed. Finally, structural fatigue analysis under a cyclic thermal load with 0.5 MW/m2 was also performed. The results indicated that there was a large discrepancy in the tile results for the two connections, while the differences in the heat sink result were not significant. The advantages and disadvantages of two connections also be discussed at last.

Microstructure and thermal stability of a structurally graded tungsten and reduced activation ferritic/martensitic steel joint

The leading design for plasma-facing high heat flux components in proposed fusion reactors involves joining plasma-facing tungsten tiles to underlying reduced activation ferritic-martensitic (RAFM) steel structures. Due to significant differences in the physical properties between W and steel, an effective method for joining them while preserving the mechanical and microstructural properties under service thermo-mechanical load is yet to be demonstrated. A transitional multilayer consisting of three layers (VCrTi, VCrAl, and FeCrAl) between W and steel is designed with the help of thermodynamic simulation and diffusion kinetics in an attempt to form solid solution bonding without forming brittle intermetallic phases. The solid solution bonding is desirable to mitigate the drastically different thermal expansion between W and steels. The multilayer structure was fabricated using spark plasma sintering (SPS). The microstructure of the bonded transition layers was analyzed after long-term annealing at 620 °C up to 1000h, through scanning electron microscope, energy dispersive spectroscopy and electron backscattering diffraction. Interdiffusion kinetics between layers and reliability of mobility database were also analyzed. This work provides a good understanding of thermal stability as well as the efficacy of multilayer functionally graded transition layer for joining W to ferritic/martensitic steel.

Performance of tungsten plasma facing components in the stellarator experiment W7-X: Recent results from the first OP2 campaign

The transition to reactor-relevant materials for the plasma facing components (PFCs) is an important and necessary step to provide a proof of principle that the stellarator concept can meet the requirements of a future fusion reactor by demonstrating high performance steady-state operation. As a first step to gain experience with tungsten as plasma-facing material in the Wendelstein 7-X (W7-X) stellarator, graphite tiles coated with an approximately 10 µm MedC tungsten layer (NILPRP Bucharest) were installed to complete the ECRH beam dump area in two of the five plasma vessel modules over an area of approximately one square meter each. In addition, tungsten baffle tiles are installed (40 tiles in total) with either bulk tungsten as part of NBI shine-through target or with a tungsten heavy alloy (W95-Ni3.5-Cu1.5) to replace the graphite tiles that were previously thermally overloaded. Based on an advanced diffusive field line tracing method and EMC3-Eirene simulations, the overloaded baffle tiles were redesigned by making the tiles thinner (i.e. moving the plasma-facing surface (PFS) away from the hot plasma region) and by reducing the local angle of incidence through toroidal displacement of the watershed. Significant erosion of the tungsten tiles can only be expected if sputtering by impurity ions such as carbon or oxygen ions contributes. However, the resulting central concentration of tungsten and the corresponding radiation losses are expected to be marginal. The expected deposition of carbon on the tungsten surfaces in the baffle regions mitigates further the possible tungsten enrichment in the core plasma. In OP2.1, no adverse effects of the installed tungsten PFCs on the plasma performance were observed during normal plasma operation. With the design changes made in the baffle area, the predicted heat load reductions could be experimentally confirmed.

Surface features of strongly heated bulk tungsten divertor plates at ASDEX Upgrade

• Horizontal cracks and humps at the surface were observed at ASDEX Upgrade solid tungsten divertor plates. • Recrystallization and grain growth is found at these locations. • High thermal stress due to temperature gradients is discussed to produce these cracks. • He bubble formation in tungsten is discusses as source of the surface humps. Solid tungsten tiles exposed at the outer strike point module of the ASDEX Upgrade divertor show cracks parallel to surface in the bulk and humps at the surface. At ASDEX Upgrade these horizontal cracks were observed for the first time. About 10 % of all tiles exhibit protrusions originated by horizontal cracks about 3-5 mm below the surface. The position of this damage coincidence with the location of highest deposited energy, which varies for the different tiles due to the mounting. The tungsten at the locations of the horizontal cracks exhibits recrystallization and grain growth, presumably due to the worse thermal contact caused by the horizontal cracks. Although, the operation of AUG was not hampered by these damages, these tiles had to be exchanged preventing progression of these damages. High thermal stress due to temperature gradients is discussed to originate these cracks. Additionally, humps are found at the surface at the position of high heat flux. Their occurrence seems to be related to the increased number of high power discharges. Focused ion beam prepared cross sections indicate pores and cavities in and below these humps. This finding is discussed in the view of He bubble formation in tungsten.

Hydrogen isotope analysis in W-tiles using fs-LIBS

Laser-Induced Breakdown Spectroscopy (LIBS) is a promising technology for in-situ analysis of Plasma-Facing Components in magnetic confinement fusion facilities. It is of major interest to monitor the hydrogen isotope retention i.e. tritium and deuterium over many operation hours to guarantee safety and availability of the future reactor. In our studies we use ultraviolet femtosecond laser pulses to analyze tungsten (W) tiles that were exposed to a deuterium plasma in the linear plasma device PSI-2, which mimics conditions at the first wall. A high-resolution spectrometer is used to detect the Balmer-alpha transition of the surface from implanted hydrogen isotopes (H and D). We use Calibration Free CF-LIBS to quantify the amount of deuterium stored in W. This proof-of-principle study shows the applicability of femtosecond lasers for the detection of low deuterium concentration as present in first wall material of prevailing fusion experiments.

Characterization of the impurity features deposited on the boronization tungsten tiles exposed in KSTAR tokamak using laser-induced breakdown spectroscopy

• The boron and other impurity elements such as: carbon, silicon, calcium and copper deposition on the surface of tungsten divertor tiles removed from KSTAR has been measured by LIBS. • A subtraction method is proposed to eliminate the W line contribution for B deposition measurement. • The lateral-distribution of the boron and carbon on the three tungsten castellated tiles (Rounded, Chamfered and Double chamfered) were carried out at 10 −5 mbar vacuum condition. Boronization is an important wall conditioning procedure for realizing the steady-state high performance operation of tokamaks such as KSTAR and ITER. The measurement of boron and its distribution on the first wall and divertor has a significant impact on studying the plasma-wall interaction (PWI) such as impurity generation, transportation and deposition behaviors. In this study, the laser-induced breakdown spectroscopic (LIBS) diagnostic method was applied in the lab for investigating the boron impurity deposited on the surface of different shaped tungsten (W) castellated tiles which were removed from the divertor region in the Korea Superconducting Tokamak Advanced Research (KSTAR). Meanwhile, other various impurity elements comprising carbon, silicon, calcium and copper were also observed. A subtraction method is proposed to eliminate the W line contribution for B deposition measurement. The depth profile of the impurity elements in the deposited layer also has been studied by the variation of spectral intensity with consecutive laser pulses. The lateral-distribution of the boron and carbon on the three tungsten castellated tiles (Rounded, Chamfered and Double chamfered) were carried out at 10 −5 mbar vacuum condition. The distribution features of the impurities on the three different shaped tungsten tiles provide valuable information to understand PWI process in KSTAR. This study demonstrates that LIBS is a potential diagnostic approach for monitoring the impurity deposition in tokamak such as KSTAR or ITER.

Conceptual design of reciprocating probes and material-testing manipulator for tokamak COMPASS Upgrade

Three new in-vessel manipulators are designed and built for the new COMPASS Upgrade tokamak with uniquely high vessel temperature (250–500 °C) and heat flux density (perpendicular to divertor surface q ⊥ ∼ 80 MW/m2 and q ∥ ∼ GW/m2 at separatrix), which challenges the edge plasma diagnostics. Here we show their detailed engineering designs supported by heat conduction and mechanical models. Deep reciprocation of electrostatic probes near the separatrix should be possible by optimizing older concepts in (a) the head and probe geometry, (b) strongly increasing the deceleration up to 100× gravity by springs and strengthening the manipulator mechanical structure. One reciprocates close to the region of edge plasma influx (the outer midplane), the other at the plasma sink (between the outer divertor strike point and X-point), for studying the plasma divertor (impurity-seeded) detachment and liquid metal vapor transport. Both probe heads are equipped with a set of ball-pen and Langmuir probes, measuring reliably and extremely fast (10−6 s) local (1 mm resolution) plasma potential, density, electron temperature and heat flux and even ion temperature with 10−5 s resolution. The divertor manipulator (without reciprocation) will place various material test targets at the outer divertor. Unique will be its capability to increase 15× the surface heat flux with respect to the surrounding tungsten tiles just by controllable surface inclination of the test targets. We plan to test liquid metal targets where such inclined surface was found critical to achieve the desired mode with lithium vapor shielding. Even in the conservative expected performance of COMPASS Upgrade, we predict to reach and survive the EU DEMO relevant heat fluxes.

Improvement of Mechanical Properties with Non-Equimolar CrNbTaVW High Entropy Alloy

CrNbTaVWx with (x = 1 and 1.7) high entropy alloys have been devised for thermal barriers between the plasma-facing tungsten tiles and the copper-based heat sink in the first wall of fusion nuclear reactors. These novel materials were prepared by ball milling and consolidated by Upgrade Field Assisted Sintering Technology at 1873 K under an applied pressure of 90 MPa for 10 min. In this work, the structural and mechanical properties of these materials were evaluated. Consolidated samples presented a major phase with a bcc-type structure with lattice parameter value of 0.316 nm for CrNbTaVW and CrNbTaVW1.7 compositions. Moreover, observation of the microstructures evidences also two minor phases: Ta-Nb-Cr and Ta-V rich (in which carbon is detected). Despite the similarity in the structural properties of these two alloys, their mechanical properties are distinct. The flexural stress for the sample with higher amount of W (CrNbTaVW1.7) is higher by 50% in the 298–873 K range, with an increased strain to fracture, which can be associated with reduced brittleness caused by the additional W incorporation.

A high efficiency fast-response gamma detector with mrad pointing capabilities

A High Efficiency fast-Response GAmma detector (HERGA) for the identification of radionuclides and for gamma-ray imaging is presented. The detector is compact, ideal for real-time analysis, with low power consumption, suitable for industrial process control, environmental monitoring and localization of nuclear wastes. A prototype consisting of 16 CsI(Tl) scintillators coupled to photomultiplier tubes (PMTs) with a digital readout was built. The scintillators are arranged in a 4 × 4 matrix; a 7 × 7 mask composed by plastic and tungsten tiles has been used to validate the imaging figures of merit according to coded mask principle. A 2D Kolmogorov–Smirnov pattern recognition statistical method has been developed, allowing to achieve a point spread function (PSF) of a few mrad for pointlike sources. The Minimum Detectable Activity (MDA) was also determined in the case of pointlike radioactive sources.

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

The divertor target components for the Chinese fusion engineering test reactor (CFETR) and the future experimental advanced superconducting tokamak (EAST) need to remove a heat flux of up to ∼20 MW m−2. In view of such a high heat flux removal requirement, this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology. Rectangular water-cooled channels with a special thermal transfer structure (TTS) are designed in the heat sink to improve the flat-tile divertor target’s heat transfer performance (HTP). The parametric design and optimization methods are applied to study the influence of the TTS variation parameters, including height (H), width (W*), thickness (T), and spacing (L), on the HTP. The research results show that the flat-tile divertor target’s HTP is sensitive to the TTS parameter changes, and the sensitivity is T > L > W* > H. The HTP first increases and then decreases with the increase of T, L, and W* and gradually increases with the increase of H. The optimal design parameters are as follows: H = 5.5 mm, W* = 25.8 mm, T = 2.2 mm, and L = 9.7 mm. The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method. A flat-tile divertor mock-up is developed according to the optimized parameters. In addition, high heat flux (HHF) tests are performed on an electron beam facility to further investigate the mock-up HTP. The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m−2 under the tungsten tile thickness <5 mm and the flow speed ≥7 m s−1. The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by ∼13% and ∼30% compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock, respectively. The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of 20 MW m−2 of heat load without visible deformation, damage, and HTP degradation. The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only ∼930 °C. The flat-tile divertor target’s HTP is greatly improved by the parametric design and optimization method, and is better than the ITER-like monoblock and the flat-tile mock-up for the WEST divertor. This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.

Tritium retention in plasma facing materials of JET ITER-Like-Wall retrieved from the vacuum vessel in 2012 (ILW1), 2014 (ILW2) and 2016 (ILW3)

ITER-Like-Wall (ILW) project has been carried out at Joint European Torus (JET) to test plasma facing materials relevant to International Thermonuclear Experimental Reactor – ITER [1]. Limiters and an upper dump plate of the vacuum vessel are made of bulk beryllium tiles, whereas for the divertor bulk tungsten and tungsten-coated carbon fibre (CFC) composite tiles are used.

In situ method for studying stresses in a pulse-heated tungsten plate based on measurements of surface curvature

Abstract Damage of divertor tungsten tiles as a result of transient plasma processes is one of the key problems of the ITER tokamak. Even if the temperature of the divertor tiles is maintained below the melting threshold, their thermal deformation, accumulation of residual stresses and cracking will occur. Cracking reduces the resistance of the metal to subsequent loads, therefore, the study of this process is important. In materials science, non-destructive methods for measuring mechanical stresses are very limited. In this work, a new method is proposed that allows non-contact investigation of stresses in polycrystalline tungsten plates by their bending. A corresponding diagnostic system and theoretical model are implemented. A number of experiments were carried out using this diagnostic system at the BETA facility at the Budker Institute to study the dynamics of tungsten plates bending with different thickness, subjected to pulsed heating of different intensities. The experimental results showed the qualitatively expected process of the plate bending, as well as good agreement with the theoretical model. This proved that the method described in the work is promising for use in order to study the residual stresses in a pulse-heated metal sample.

Design and manufacturing of bulk tungsten tiles for WEST outer limiter

Limiter objects are facing high loads, even if they are not as close to the plasma as divertor for example. Loads come from plasma interaction during the standard situation (heat loads) and during abnormal situation like disruption (electromagnetic loads, runaway electrons). In WEST, the initial guess was to use CFC tiles with a tungsten coating for the outer bumper limiter. After two campaigns, it appears that the power and quantity of runaway beams had been underestimated, leading to early damages of the coating on significant surfaces. A second attempt for the tiles is to use bulk tungsten. This should ensure a high resistance the runaway beams. The design is validated for the given specifications. The tiles are manufactured in early 2019, and used during the WEST operations later this year. The first observations show a good resistance to the runaway beam, as expected.

The bubble problem of the plasma facing material: A finite element study

The damage of first wall material in fusion reactor due to the bubbles caused by plasma has been studied by introducing a relation between the von Mises equivalent stress and the temperature field. The locations and shapes of the bubbles and the synergetic effect between the different bubbles under steady operational conditions have been studied using the finite elements method. Under transient heat loads, plastic deformations have been found to occur, and are significantly enhanced by the presence of the bubbles. The calculated concentration locations of von Mises equivalent stress are well consistent with the observed crack positions of the tungsten surface in many test experiments. Our simulations show that the damage of the bubbles is not severe enough to lead to catastrophic failure of the tungsten armor; however, it can cause local and gradual detachment of tungsten surface, which provides a reasonable explanation for the observed pits and rough or hairy surface morphology etc. Considering the transient heat loads, the lower bound of the security thickness of the tungsten tile is estimated to be greater than 2 mm.