First Wall Research Articles

Differential cross section measurements of the 7Li(3He, p0)9Be, 7Li(3He, p2)9Be, and 7Li(3He, d0)8Be reactions from 2 to 4 MeV

The use of lithium in plasma facing components (PFCs) for fusion reactors is increasingly more prominent due to its ability to enhance plasma performance and potential to protect the underlying high-Z PFCs. In order to use nuclear reaction analysis (NRA) to investigate lithium depth profiles in fusion relevant experiments such as hydrogen retention in lithiated PFCs, there is a need for experimental measurements of the cross section of the reactions of interest over a wide range of energies and in a scattering angle in which the differential cross sections of other relevant species, such as deuterium and oxygen, have been measured. In this paper we report the differential cross sections of the 7Li(3He, p0)9Be, 7Li(3He, p2)9Be, and 7Li(3He, d0)8Be reactions from 2 to 3.919 MeV at a measurement angle of 135°. The measurements were performed on a lithium fluoride film target with the 7Li areal density determined using the 7Li(p, α)4He reaction at 2 MeV and 150°.

Synthetic measurements of runaway electron synchrotron emission in the SPARC tokamak.

With plasma currents up to 8.7 MA, the SPARC tokamak runs the risk of forming multi-MA beams of relativistic "runaway" electrons (REs), which could damage plasma facing components if unmitigated. The infrared (IR) and visible imaging and visible spectroscopy systems in SPARC are designed with measurements of synchrotron emission from REs in mind. Synchrotron radiation is emitted by REs along their direction of motion, opposite the plasma current. Matched clockwise and counterclockwise wide views are proposed to detect synchrotron and background radiation, allowing observation of RE synchrotron emission in both plasma current configurations. Due to SPARC's high toroidal magnetic field strength, 12.2T on axis, the synchrotron light spectrum is expected to peak in the visible-IR wavelength range. The synthetic diagnostic tool, Synchrotron Orbit-Following Toolkit, is used to model synchrotron images and spectra for three scenarios, with appropriate magnetic equilibria for each: REs generated during plasma current ramp-up, steady-state flat-top (although unlikely, but serving as a reference), and disruptions. Required time resolutions, achievable spatial coverage, and appropriate spectral ranges for various RE energies are assessed.

Design, justification, and prototyping of the visible and infrared wide angle viewing system diagnostic for ITER equatorial port 12.

The ITER equatorial visible and infrared Wide Angle Viewing System (WAVS) will play a major role in the protection of plasma facing components by providing surface temperature measurements of these components. It will also image the plasma emission in the visible range. The WAVS is composed of 15 lines of sight located in four Equatorial Ports (EPs) 3, 9, 12, and 17. Its development is being carried out by the Consortium constituted by CEA, CIEMAT, INTA, and Bertin Technologies, within grant contracts financed by F4E. In the EP12, the in-vessel and ex-vessel components of the WAVS are at their final design phase. This article presents an overview of the opto-mechanical design of the WAVS in the EP12 presented at the final design review.

The infrared thermography system on the MAST-U tokamak

Power loading from plasma in the scrape-off layer limits the lifetime of plasma-facing components in tokamak-based power plants. The Mega Ampere Spherical Tokamak Upgrade [W. Morris et al., IEEE Trans. Plasma Sci. 46(5), 1217–1226 (2018)] (MAST-U) features four divertor strike points (SP) owing to its up-down symmetry. This paper introduces the five-camera infrared thermography system, covering all four SPs and their mode of operation on MAST-U. The system employs both medium and long wavelength cameras, offering high frame rates up to 4 kHz and spatial resolutions of down to 3 mm. Postprocessing involves artifact elimination and heat flux calculation using in-house analysis software that incorporates the inverse heat transfer code, THEODOR (THermal Energy Onto DivertOR).

The effect of CD4 injection on WD molecule sputtering at the divertor target in EAST

Tungsten (W) is one of the preferred plasma-facing materials for future fusion reactors because of the advantages of high melting point, low physical sputtering rate and low fuel retention. The lifetime and service performance of W materials are limited by erosion processes. In recent years, tungsten deuteride molecular (WD) sputtering has been discovered in tokamak. Studying WD sputtering helps to better understand the erosion of tungsten materials in tokamak. EAST adopts the method of injecting impurity gas in the divertor to alleviate the particle flux and heat flux bombarding the tungsten divertor, and CD4 is one of the impurities used. CD4 injection experiments on EAST show that after CD4 is injected from the upper outer (UO) divertor target, despite that the electron temperature (Te) around the strike point on the UO divertor target drops to about 8 eV, strong WD sputtering occurs in the far scrape-off layer (SOL) region. The divertor probe data shows that after CD4 injection, the particle flux from the upstream plasma onto the divertor split into two bands, creating a new particle flux strike point in the far SOL region. The energy of incident particles at the original strike point is significantly reduced, while the energy of incident particles at the far SOL region is enhanced. The edge electron density profile at mid-plane indicates that CD4 injection causes an increase of density in the SOL, which could improve the coupling of Lower-Hybrid Wave (LHW) with the edge plasma as evidenced by the decreasing reflected power of LHW. The increase in absorbed power of LHW could change the particle and energy transport in the edge plasma, thus plays an important role in the enhanced WD sputtering in the far SOL region after CD4 injection.

Comprehensive Screening of Plasma-Facing Materials for Nuclear Fusion

Plasma-facing materials (PFMs) represent one of the most significant challenges for the design of future nuclear fusion reactors. Inside the reactor, the divertor will experience the harshest material environment: intense bombardment of neutrons and plasma particles coupled with extremely large and intermittent heat fluxes. The material designated to cover this role in ITER is tungsten. While no other materials have shown the potential to match the properties of tungsten, many drawbacks associated with its application remain, including cracking and erosion induced by a low recrystallization temperature combined with a high ductile-brittle transition temperature and neutron-initiated embrittlement; surface morphology changes (helium bubbles and fuzz layer) due to plasma-tungsten interaction with subsequent risk of spontaneous material melting and delamination; and low oxidation resistance in the case of air contamination. Exploring alternatives to tungsten requires the design of a multivariable optimization problem. This work aims to produce a structured and comprehensive material screening of PFM candidates based on known inorganic materials. The method applied in this study to identify the most promising PFM candidates combines peer-reviewed data present in the PAULING FILE database and first-principles density-function theory calculations focusing on two key PFM defects—namely, the sputtering of surface atoms and the incorporation of interstitial hydrogen, respectively characterized by the surface binding energy and the interstitial formation energy. The crystal structures and their related properties, extracted from the PAULING FILE, are ranked according to the heat-balance equation of a PFM subjected to the heat loads in the divertor region of an ITER-like tokamak. The materials satisfying the requirements are critically compared with the state-of-the-art in plasma-facing materials research and in studies of refractory materials exposed to high temperatures, plasma, and neutron bombardment. This comparison assesses their thermo-mechanical properties under such conditions, identifying a subset of promising materials for first-principles electronic structure calculations. Most previously known and extensively studied PFMs, such as tungsten, molybdenum, and carbon-based materials, are captured by this screening process, confirming its reliability. Additionally, less familiar refractory materials suggest performance that calls for further investigations. Published by the American Physical Society 2024

Gyrokinetic prediction of core tungsten peaking in a WEST plasma with nitrogen impurities

Abstract Tungsten peaking is predicted in the core of a WEST plasma with total-f gyrokinetic simulations, including both collisional and turbulent transport. This prediction is validated with a synthetic diagnostic of the bolometry. Although nitrogen impurities are shown to reduce the neoclassical peaking of tungsten on-axis, the overall tungsten peaking increases when nitrogen impurities are present, as they reduce the turbulence screening off-axis. This finding is important for the plasma current ramp-up phase of ITER, where light impurities seeding will be desirable to achieve low temperatures at the plasma-facing components and reduce tungsten sputtering. It provides further argument for applying early ECRH heating to maintain margins on the core power balance. The neoclassical peaking factor is cross-verified between XGC and FACIT. The heat flux at separatrix and the heat load width are modeled by XGC and compared to WEST data.

Identification of a double decay length ([formula omitted]) heat flux deposition shape with embedded thermal measurement and neural network

The estimation of the heat flux density distribution profiles in tokamak devices is a very important research topic for edge plasma physics purposes and also to ensure the safety of the machine. In the radial direction, the heat flux exhibits an exponential decay that could be captured by thermal sensors distributed in the plasma facing components. Radially distributed thermal sensors based on Fiber Bragg grating technology have been embedded in the WEST lower divertor to study the heat flux deposition profiles during plasma operation. The comparison between embedded measurements and a 3D finite element model shows a small decay length (5 – 10 mm) on top of a wider heat flux with a decay length around 30 to 50 mm. A tool using neural network has been developed in order to predict the values of the different parameters describing the deposited heat flux from embedded temperature measurements in steady state. A large span of deposited heat fluxes with maximum heat flux ranging from 1 to 9 MW/m2 and decay length from 5 to 50 mm were characterized using this tool over a database of more than 250 experimental L-mode pulses performed in WEST in attached divertor configuration. The comparison of the predicted heat flux parameters values with macroscopic plasma parameters have revealed the appearance of the narrow component with the increase of the divertor power load (Pdiv) with a threshold dependant of the plasma current (IP).

Research on the characteristics of the Helium plasma beam in HIT-PSI

As a high heat flux linear plasma device designed for studying divertor materials in future fusion reactors, HIT-PSI(Plasma Surface Interaction device at Harbin Institute of Technology) has been successfully constructed and has maintained stable operation since its completion. The characteristics of He plasma beams in HIT-PSI are investigated by emission spectroscopy and an infrared camera, with preliminary irradiation experiments conducted by bombarding tungsten with the beam. For relatively conservative discharge parameters, HIT-PSI achieved a steady-state heat flux capacity of ∼40 MW/m2 using infrared measurements, with the full width at half maximum (FWHM) of the heat flux beam reaching 4 mm. These characteristics make HIT-PSI an advanced platform for testing divertor materials and plasma-facing components, providing essential experimental supports for research and development of high-performance divertor materials.

A First-Principles Study of the Structural and Thermo-Mechanical Properties of Tungsten-Based Plasma-Facing Materials

Tungsten (W) and tungsten alloys are being considered as leading candidates for structural and functional materials in future fusion energy devices. The most attractive properties of tungsten for the design of magnetic and inertial fusion energy reactors are its high melting point, high thermal conductivity, low sputtering yield, and low long-term disposal radioactive footprint. Despite these relevant features, there is a lack of understanding of how the structural and mechanical properties of W-based alloys are affected by the temperature in fusion power plants. In this work, we present a study on the thermo-mechanical properties of five W-based plasma-facing materials. First-principles density functional theory (DFT) calculations are combined with the quasi-harmonic approximation (QHA) theory to investigate the electronic, structural, mechanical, and thermal properties of these W-based alloys as a function of temperature. The coefficient of thermal expansion, temperature-dependent elastic constants, and several elastic parameters, including bulk and Young’s modulus, are calculated. Our work advances the understanding of the structural and thermo-mechanical behavior of W-based materials, thus providing insights into the design and selection of candidate plasma-facing materials in fusion energy devices.

Turbulent transport mechanisms and their impact on the pedestal top of JET plasmas with small-ELMs

Abstract Understanding the transport mechanisms that regulate the H-mode pedestal structure is crucial for developing regimes with edge dynamics that are compatible with the thermal and particle load constraints of plasma facing components. In this regard, at the JET tokamak, H-mode plasma regimes featuring small-ELM (BSE) dynamics have been developed, expelling less plasma energy per ELM compared to the standard type-I ELM. The pedestal of these new regimes remains stable against peeling-ballooning modes, leaving its structure unexplained.
In this study, three baseline H-mode JET discharges, a reference with type-I ELM and two BSE regimes, are selected to investigate the turbulent transport mechanisms occurring in pedestals with different structures. The turbulence analysis is conducted using local gyrokinetic simulations around the pedestal top position. The results show that at ion-scales, while hybrid ITG-KBM and KBM are destabilized in the type-I ELM discharge, the BSE regimes are dominated by hybrid ITG-TEM, suggesting a different role of turbulent transport in the two cases. At electron-scales, both regimes are dominated by toroidal and slab ETG, the latter extending up to very small scales. Simulations of turbulence driven by hybrid ITG-TEM focuses on the impact and competition of electromagnetic (EM) effects and equilibrium $\vb*{E}\times\vb*{B}$ shearing. A strong EM stabilization mechanism is found, resulting in a saturated turbulence regime with spectra different from those in the electrostatic limit (ES$_l$). Additionally, while increasing $\vb*{E}\times\vb*{B}$ shearing reduces transport in ES$_l$ simulations, it is observed to have the opposite effect in the EM regime. Finally, stiff ETG transport is also shown to produce transport levels compatible with experimental observations.
The findings of this study suggest the EM stabilization and its competition with $\vb*{E}\times\vb*{B}$ shearing as new key elements determining the nature of the turbulent transport mechanisms around the pedestal top, which need to be considered in reduced models for pedestal limits.

Power handling in a highly-radiative negative triangularity pilot plant

Abstract This work explores detailed power handling solutions for a class of high-field, highly-radiative negative triangularity (NT) reactors based around the MANTA concept [Rutherford et al. (2024)]. The divertor design is kept as simple as possible, opting for a standard divertor with standard leg length. FreeGS is used to create an equilibrium for the boundary region, prioritizing a short outer leg length of only ∼50 cm (∼40% of the minor radius). The UEDGE code package is used for the boundary plasma solution, to track plasma temperatures and fluxes to the divertor targets. It is found that for PSOL = 25 MW and nsep = 0.96 × 1020 m−3, conditions consistent with initial core transport modeling, little additional power mitigation is necessary. For a fixed impurity fraction of just 0.13\% Ne in the plasma, the peak heat flux density at the more heavily loaded outer targets falls to 7.8 MW/m2, while the electron temperature, Te, remains just under 5 eV. Scans around the parameter space reveal that even at densities lower than in the primary operating scenario, PSOL can be increased up to 50 MW, so long as a slightly higher fraction of extrinsic radiator is used. With less than 1% neon (Ne) impurity content, the divertor still experiences less than 10 MW/m2 at the outer target. Design of the plasma-facing components includes a close-fitting vacuum vessel with a tungsten inner surface as well as FLiBe-carrying cooling channels fashioned into the VV wall directly behind the divertor targets. For the seeded heat flux profile, Ansys Fluent heat transfer simulations estimate that the outer target temperature remains at just below 1550◦C. Initial scoping of advanced divertor designs shows that for an X-divertor, detachment of the outer target becomes much simpler, and plasma fluxes to the targets drop considerably with only 0.01% Ne content.

Divertor erosion at ASDEX Upgrade during helium plasma operations

The effect of helium plasma operation on the erosion of plasma-facing components at the low-field side divertor of ASDEX Upgrade was investigated during the 2022 helium experimental campaign. A set of tungsten-covered graphite samples with small platinum marker spots was exposed to both L-mode and H-mode plasma discharges. The highest net erosion of over 1.1 nm/s was observed around the H-mode strike point similar to the case in deuterium plasma. Significant helium inventories of about 6 × 1016 He/cm2 were measured in the scrape-off layer region of the divertor. Impurity deposition including boron and deuterium showed a distinct peak up to 2.4 × 1017 B/cm2 and 1.0 × 1016 D/cm2 between the strike points, and significant boron inventories up to 5.9 × 1016 B/cm2 were also measured on the scrape-off layer side of the H-mode strike point. Platinum re-deposition was not detected between the marker spots, suggesting that it occurs only very locally within the markers. Overall erosion was, as expected, higher than in deuterium discharges, and it also remained comparatively high towards the scrape-off layer, unlike with deuterium.

Influence of temperature gradients on helium diffusion and clustering in tungsten: A molecular dynamics study

Tungsten and its alloys, used as plasma-facing materials in fusion reactors, endure high-flux, low-energy helium ion irradiation and temperature gradients induced by thermal load and shocks. This study utilizes molecular dynamics (MD) simulations to explore the diffusion and clustering behavior of helium (He) in tungsten (W) under temperature gradients. The migration behavior of isolated helium atoms, small helium clusters, and helium self-interstitial atom (He-SIA) clusters within tungsten is analyzed. It is revealed that both helium and He-SIA clusters tend to migrate towards higher temperature regions, exhibiting negative thermophoresis. The nucleation of helium clusters, the formation of Frenkel pairs, and the interactions between self-interstitial clusters and helium clusters are also investigated. The results indicate that helium concentration significantly impacts He nucleation behavior. Directional diffusion may lead to areas of high concentration over time, potentially promoting the formation of He clusters and bubbles. These insights enhance the understanding of tungsten's performance and durability as a plasma-facing component in fusion reactors.

Plasma edge and scrape-off layer turbulence in gyrokinetic simulations of negative triangularity plasmas

Abstract Gyrokinetic simulations in the long-wavelength or drift-kinetic limit are carried out of DIII-D inner-wall-limited (IWL) plasmas to investigate the effect of triangularity on edge and scrape-off layer (SOL) turbulence. The effect of neutral interactions and triangularity on plasma blobs is explored due to the impact blobs can have in setting the SOL width or introducing impurities through interactions with plasma-facing components. Seeded blob simulations with neutrals in shaped SOL scenarios demonstrate that increasing elongation, triangularity, or Shafranov shift decreases radial blob velocities, but neutral interactions have a minor effect. Fully turbulent simulations of DIII-D IWL plasmas include both open- and closed-field-line regions. The negative triangularity (NT) simulation has lower average core Te , lower normalized Te fluctuations, and lower fluxes, but a greater number of coherent structures (blobs) identified with increased size and velocity, on average. Density and electron temperature profiles are within a factor of 2 of experimental values. The increased trapped electron particle fraction in NT simulations is consistent with previous studies.

Thermal properties of MB2-WC (M = Ti, Zr, Hf) and tungsten and their stability after deuterium plasma exposure

The thermal properties of ultra-high temperature ceramics (UHTCs) in the MB2-WC (M = Ti, Zr, Hf) system and tungsten were studied for potential application as plasma-facing materials in fusion power plants. The sintered UHTC and tungsten samples were subjected to deuterium plasma or protons irradiation. Thermal diffusivity was measured using the laser flash method, and superficial thermal conductivity was analyzed through atomic force microscopy. Results showed that the thermal properties did not degrade when exposed to relevant environments and remained stable over a range of temperatures, unlike the reference tungsten material. Thermal conductivity ranged from 61 to 68 W m−1 K−1 for TiB2-2(WC-6Co), from 53 to 63 W m−1 K−1 for ZrB2-6WC, from 67 to 75 W m−1 K−1 for HfB2-6WC, and from 180 to 119 W m−1 K−1 for tungsten across the temperature range from room temperature to 1200 °C. The increasing trend of thermal effusivity, over 19000 J s−0.5 m−2 K−1 at 1200 °C, justifies further testing and of UHTC materials for fusion applications.

First integrated core-edge fluid simulation of ITER’s Limiter–Divertor transition with SolEdge-HDG

This work explores the Limiter–Divertor transition (L–D) during the current ramp-up of ITER’s Q = 10 baseline plasma scenario at various central line-integrated density nli values. The analysis, based on transport simulations performed with the latest version of SoleEdge-HDG, focuses on the time evolution of heat and ion particle fluxes, revealing regions of elevated temperature on the inner wall and plasma-facing components (PFCs) despite moderate loads. The investigation also delves into the effects of perpendicular convection flux terms on density build-up, comparing different formulations and their interplay with auxiliary heating sources. Furthermore, the paper shows the impact of taking into account the evolution of the parallel neutral momentum on plasma and neutral density at the targets in the context of an ITER steady-state scenario.

Development of ITER first wall heat load feedback control

Real-time monitoring and control of thermal loads on tungsten plasma-facing components is mandatory for ITER to avoid their overheating during plasma discharges. For the Start of Research Operations (SRO) phase, dedicated First Wall Heat Load Control (FWHLC) functions are included in the ITER Plasma Control System (PCS) to prevent the development of excessive plasma heat loads on the inertially cooled first wall (FW). In this work, we propose a FWHLC design with a model-based approach, which features a simplified thermal model for the FW armour. This control-oriented model simulates the thermal response of ITER FW to slow changes in plasma equilibrium and energy, which during ITER operation will be monitored by the real-time wide angle infrared camera system. This allows computationally efficient runs for rapid controller performance assessment but can also assist operation scenario design by pre-empting potential heat load issues. FWHLC is designed to react to measured and predicted FW temperatures overcoming predesigned limits with real-time variations of the plasma shape or auxiliary power references, aiming to reduce thermal loads before a premature plasma termination is required to protect the FW. The new scheme is tested in closed loop simulations, where perturbations to nominal ITER low current scenarios are introduced in the wall thermal model to trigger the response of FWHLC, supporting the effectiveness of the proposed policy.

Toroidal distribution of heat load on castellated plasma facing components for lower divertor target in EAST

In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

First SOLEDGE3X-EIRENE simulations of the ITER Neon seeded burning plasma boundary up to the first wall

Boundary plasma simulations are essential to estimate expected divertor and first wall (FW) heat and particle loads on ITER during burning plasma operation. A key missing feature of existing SOLPS simulations (Pitts et al., 2019) is the absence of a plasma solution out to the main chamber walls, essential to self-consistently estimate the gross sputtering of wall material. Here, SOLEDGE3X is applied for the first time to obtain up-to-the wall burning plasma solutions of the ITER boundary plasma at the nominal PSOL = 100 MW of the main SOLPS database simulations, including He ash, Ne seeding but without fluid drifts. Compared with the most recent SOLPS-ITER simulations, our simulations show differences in the exact impurity distribution, but the key results for divertor and wall heat flux remain consistent. In the context of the ITER re-baselining exercise (Pitts, 2024), in which the Be FW armour is proposed to be exchanged for tungsten (W), estimates of W wall sources are key to the assessment of likely core contamination and hence impact on fusion gain. We compare the W gross erosion rates due to the different species excluding W self-sputtering. For the cases simulated spanning 0.27%–0.47% separatrix-averaged Ne concentration and 7.5×1022s−1−1.95×1023s−1 D fuelling, Ne8+ remains the largest contributor to the sputtering flux with the largest source being the outer divertor and baffle. The species-wise contribution to W sputtering changes with fuelling with sputtering due to lower Ne charge states being significant at low D fuelling. In general, the gross W sputtering source is found to decrease with increase in D fuelling and increase with increased Ne seeding.