First Wall Research Articles

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.

Recent progress in the development of liquid metal plasma facing components for magnetic fusion devices

One of the most critical challenges for future fusion reactors is to develop longevity plasma-facing components (PFCs) exposed to extremely high heat and neutron loads. As opposed to those employing solid metals, PFCs with flowing liquid metals (LM) have shown self-healing, heat removal and good impurity control capabilities, all essential to fusion devices. Recently, significant progress in LM-PFC development has been reported globally, with data from several magnetic fusion devices. These studies reveal that LM-PFCs can endure extreme heat fluxes while maintaining plasma compatibility. New design concepts have been proposed and numerically analyzed, advancing models for liquid PFCs in future reactors. Despite existing technical challenges, these developments suggest that LM-PFCs hold promise for future fusion applications.

Repair of heat load damaged plasma–facing material using the wire-based laser metal deposition process

Due to its unique properties tungsten is a promising candidate as plasma-facing-material (PFM) in future nuclear fusion reactors. Tungsten features an exceptionally high melting point, high thermal conductivity, low tritium inventory and comparatively low erosion rate under plasma loading [1]. But given the extreme loads on the PFM during operation of a fusion reactor, the lifetime of plasma-facing components (PFC)s is limited. Currently, it is planned to replace damaged PFCs when they reach the end of their service life. However, the lifetime of PFCs could be increased by in situ repair using additive manufacturing technology (AM) in the form of direct-energy-deposition (DED). The wire–based laser metal deposition process (LMD-w) meets several necessary conditions for operation in the vessel and could be used for performing such in situ repairs.It was investigated if the LMD-w process is able to heal thermal induced surface cracks and roughening by remelting the substrate during deposition of tungsten. For this purpose, tungsten samples of 12 × 12 × 5mm3, which later served as substrate plates for the LMD-w experiments, were treated with combined steady-state and transient thermal loads in the electron beam facility JUDITH 2. These samples were brazed to a copper cooling structure and exposed to 105 thermal shocks of 0.5 ms duration and an intensity of Labs = 0.55 GW m−2 (FHF = 12 MW s0,5 m−2) at a base temperature of Tbase = 700 °C. This way, edge localized mode (ELM) like thermal load damage was induced on the tungsten samples. On these samples, different LMD-w and laser remelting process strategies were performed. Subsequently, these samples were analyzed, and it was examined that the healing of the pre-damaged substrate material was successful. In parallel, the laser remelting process was modeled in a thermal transient finite element method(FEM) simulation in order to gain an insight into the temperatures prevailing in the material during the process.

Designing tungsten armoured plasma facing components to pulsed heat loads in magnetic fusion machines

A possible design rule for preventing surface damage from thermal transients to solid tungsten armour is proposed and formulated for the plasma facing components (divertor, first wall) of magnetic fusion machines. The rule is based on combined results from laboratory experiments and operating fusion machines, and fundamental engineering principles such as the heat flux factor (FHF) and fatigue usage fraction (FUF). As an example, the rule would allow 2.104 transient heat loads cycles at a FHF of 10MJm-2s-½ before the lifetime is considered exhausted. The formulation of the rule using engineering principles allows combining loads of different magnitudes and various number of cycles. A practical example of the rule usage is provided, illustrating loads combination and how the rule may contribute to the component geometrical design. The proposed rule is only valid for surface loading conditions, hence is not usable for volumetric loading conditions such as runaway electrons. Setting a budget lifetime and a design rule does not preclude actual plasma operation beyond the design lifetime. It is actually normal that experimental devices explore a larger domain than the one defined at the time of the design.

Inferring the scrape-off layer heat flux width in a divertor with a low degree of axisymmetry

Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MWm−2. The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (≥1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (≥0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.

Survey of tungsten gross erosion from main plasma facing components in WEST during a L-mode high fluence campaign

An initial high fluence campaign was performed in WEST, in 2023, on the newly installed actively cooled tungsten divertor composed of ITER-grade monoblocks. The campaign consisted in the repetition of a 60 s long Deuterium L-mode pulse in attached divertor conditions, cumulating over 10000s of plasma exposure. A maximum deuterium fluence of approximately 5⋅1026m−2 was reached in the outer strike point region, representative of a few high performance ITER pulses. Gross tungsten erosion inferred from visible spectroscopy shows that the most eroded plasma facing component is the inner divertor target with rates ten times larger than on the outer divertor target. The outer midplane tungsten bumpers, located a few centimeters from the plasma, show gross erosion rates two times lower than at the outer divertor. We conclude that the outer midplane bumpers have a negligible contribution to the long range tungsten migration and deposition onto the lower divertor. The cumulated gross erosion rate on the inner divertor translates in an effective gross erosion thickness of about 20μm, while it is about 2μm for the outer divertor. Strikingly, these orderings coincide with the thickness of deposits found locally on the divertor : the exposed surfaces of high field side monoblocks are covered with several tens of μm tungsten deposits, while on the lower field side, few μm thin tungsten deposits are only found on the magnetically shadowed parts of monoblocks. The strong impact of those deposits on WEST operation, namely perturbation of surface temperature measurement with infra-red thermography, and the emission of flakes causing radiative perturbation of the confined plasma, calls for anticipating similar issues in ITER. In particular, the start of research operation shall consider the definition of a divertor erosion budget in order to anticipate the formation of deleterious deposits.

DFT Simulations of the Self-healing Behavior of a W〈110〉/W〈112〉 Grain Boundary in the Presence of Coexisting Point Defects

Light impurity atoms (LIAs), such as hydrogen and helium, tend to aggregate at pre-existing intrinsic point defects. This aggregation leads to detrimental effects, particularly in environments such as those foreseen in nuclear fusion reactors. There, such impurities would be ubiquitous, resulting in unacceptable material behavior that would unqualify the material as a Plasma Facing Material (PFM). One option to delay the degradation in performance is the use of nanostructured tungsten (NW), showing a large density of grain boundaries (GBs). Although we have already addressed the behavior of a single LIA in a GB, in this work we present the combined synergistic effects of the simultaneous presence of multiple LIAs, vacancies and Self-Interstitial Atoms (SIA) at semicoherent W/W interfaces using ab initio methods. Our results reveal a complex and interesting process in the competition between LIAs and SIAs. When the number of SIAs is low, He appears to hinder their recombination with vacancies, therefore casting doubts on the self-healing provided by NW. However, in the presence of larger numbers of SIAs, their mutual repulsion leads to the opposite behavior. Thus, a thorough thermodynamic assessment in which the evolution of the system may be tracked emerges as the crucial subsequent step in these investigations.

Calculation of in-situ steady-state heat flux on EAST lower divertor

Heat flux is a key issue in tokamak devices. The non-uniform high heat flux on Plasma-Facing Components (PFCs) has led to local severe damage, including cracks and melting, in current tokamaks such as EAST and WEST. To characterize the non-uniform heat flux loading on the divertor surfaces, the parallel incident heat flux q‖, the decay length λq along the radial direction and the Gaussian spreading width S are used. The q‖ can lead to a very high peak heat flux loading on the divertor surfaces, which may cause critical heat flux problems. Additionally, the decay length is a key consideration for future tokamak designs like ITER. Every effort on the present tokamak devices contributes to updating the scaling of the heat flux. In EAST, a calculation method based on a high spatial resolution IR camera is employed to obtain the heat flux and decay length. The main process involves comparing the surface temperature distribution calculated by Fluent simulation with that measured by an infrared camera. Taking a high heating source discharge (#123059 ∼ 10 MW heating source) as an example, the heat flux is as follows: q‖=216-14+19 MW/m2, with λq=6.2-1.1+1 mm, and S=1.2±0.4 mm; it is in line with Langmuir probe data. The infrared-based heat flux calculation method can calculate the peak incident heat flux and the decay length simultaneously, its result can help to update the scaling model of heat flux, thus not only helping to improve the present device but also offering important reference for future tokamaks