Tokamak Device Research Articles

Efficient Numerical Solution of Coupled Axisymmetric Plasma Equilibrium and Eddy Current Problems

This paper presents an efficient numerical tool for solving the problem of transient nonlinear evolution of plasma equilibrium in presence of eddy currents. The code, named FRIDA-TD (FRIDA-Time Domain), and based on the existing FRIDA (FRee-boundary Integro-Differential Axisymmetric) code, exploits two different approaches for solving the elated nonlinear system of Partial Differential Equations (PDE): (i) Finite Element Method - Boundary Element Method (FEM-BEM) for plasma equilibrium, and (ii) Axisymmetric Volume Integral (AVI) formulation for eddy currents. The resulting FRIDA-TD is a fast and flexible tool suitable for engineering-oriented purposes, such as the design of Tokamak devices, set up of plasma operations, prediction of performance scenarios, and design of feedback control systems. FRIDA-TD is successfully validated at first with the well-established CarMa0NL code, and then against experimental data of the RFX-mod device.

Simulation of effect of thermionic emission on magnetized sheath near target plate of tungsten divertor

The high confinement mode (H-mode) is a preferred operation mode of tokamak devices in the future, but the burst of edge localized mode (ELM) will sharply increase the heat load deposited on the divertor target, raising the target temperature rapidly and strengthening surface thermionic emission. In this paper, a one-dimensional fluid model is used to simulate the influence of thermionic emission on the characteristics of the magnetized sheath. The results show that the amplitude of float potential and the electric field strength both decrease under the action of thermionic emission. Plenty of thermionic emission electrons leave the target, resulting in a region with negative charge density near the target plate, and the magnetized sheath is divided into two parts: ion sheath and electron sheath. In the electron sheath, with the rise of the target surface temperature, electrons accumulated in front of the target also increase, the potential distribution is non-monotonic, and a “virtual cathode” structure appears. The reverse electric field formed near the target will confine the thermionic emission electrons leaving the target and slow down the ion movement, leading to a decrease of the ion energy deposited on the target. With the increase of the angle between the magnetic field and the target normal, the potential of the magnetized sheath and the proportion of the thickness of the electron sheath in the magnetized sheath both increase. The virtual cathode potential decreases, the temperature of the target required to form the virtual cathode rises.

Dust explosion in fusion reactors: Explosion characteristics and reaction mechanism of tungsten micro-powder

The explosion of in-vessel dusts, mainly produced by energetic plasma and wall interaction, will pose a great threat to the safe operation of a fusion reactor. Unlike dust explosions in traditional industrial systems, dust explosion in tokamak devices has the unique characteristics, including dust composition (tungsten is one of the main components), initial conditions of explosion, etc. However, the experimental data on this aspect are very few, which is not enough to support the safety assessment of fusion reactors. In this work, the experiments were used to study the explosion characteristics and mechanism of the micro-tungsten in a 20 L spherical vessel. Results showed that the lower explosion limit (LEL) of 2 µm tungsten dust was in the range of 250–500 g/m3, and the maximum explosion pressure would reach 0.46 MPa at 3000 g/m3. The factors related to fusion reactor environment, including initial pressure, ignition energies and the ignition delay time, were studied to demonstrate a significant effect on explosion parameters. The explosion product changed from WO3 to WO2.72 with the increase of nominal dust concentration. A combustion mechanism for single mirco-tungsten particle was established. Based on the experimental data, the explosion possibility and hazard of tungsten dust in fusion reactor were also dicussed.

Experimental study on plasma generated by a tapered coaxial accelerator and its damage effects on a tungsten target at different angles

Plasma wall interaction inevitably occurs during the operation of tokamaks. The coaxial gun device has low operation cost and the parameters of plasma produced by the gun are close to those of type I edge localized mode (ELM); therefore, the coaxial gun is suitable in simulation experiments as a heat flux source of transient events such as type I ELM under the condition of H-mode in the International Thermonuclear Experimental Reactor. In this paper, the plasma generated by the discharge of a tapered coaxial accelerator thermal shock on a tungsten target is used to simulate the damage effect of the divertor. The plasma parameters are measured in the experiment. The velocity of the plasma is 41.7 km s−1, and the kinetic energy of a single hydrogen ion is 9.2 eV. The energy density at the center of the plasma can reach 1.5 MJ m−2, and the density can reach about 2.78 × 1015 cm−3. The reflection of plasma in the process of exposure at different angles is observed. It is observed that droplets of millimeter size splash from the target. Traces of liquid flow are observed on the surface of the target, which shows that there is a melting process on the surface of the target. The mass loss of the target is of the order of milligrams after 20 pulses. The ablation and residual stress of the target surface both decrease with a decrease in the angle. This is because the accumulated energy per unit area of the target surface decreases with a decrease in the angle. The results of the simulation experiment help us to understand the working state around the divertor target in tokamak devices.

Triangularity effects on global flux-driven gyrokinetic simulations

On the road to fusion energy production, many alternative scenarios have been investigated in order to address certain well-known problems of tokamak devices; among which, anomalous transport, ELMs and disruptions. The studies on plasma shaping fall into this effort.In particular, it has been experimentally observed that when operating in L mode, negative triangularity (NT) features better confinement properties than positive triangularity (PT). However, even though the trend is quite clear, a complete and satisfying theoretical explanation for this experimental findings is still lacking.With the aim of understanding and describing these improvements starting from first principles, we present the first comparison between PT and NT with global flux-driven gyrokinetic simulations performed with the ORB5 code.The numerical setup includes: electrostatic turbulence, kinetic trapped electrons, non-linear collisional operator, ECRH source, limiter and wall as boundary conditions. The simulations have been performed on ideal MHD equilibria and kinetic profiles inspired by TCV experiments, in a mixed ITG-TEM regime.First analysis reveal a strong reduction of transport in NT; while at the edge PT shows superdiffusivity, NT does not. The limiter plays an important role that has to be further clarified.

Surrogate models for plasma displacement and current in 3D perturbed magnetohydrodynamic equilibria in tokamaks

A numerical database of over one thousand perturbed three-dimensional (3D) equilibria has been generated, constructed based on the MARS-F (Liu et al 2000 Phys. Plasmas 7 3681) computed plasma response to the externally applied 3D field sources in multiple tokamak devices. Perturbed 3D equilibria with the n = 1–4 (n is the toroidal mode number) toroidal periodicity are computed. Surrogate models are created for the computed perturbed 3D equilibrium utilizing model order reduction (MOR) techniques. In particular, retaining the first few eigenstates from the singular value decomposition (SVD) of the data is found to produce reasonably accurate MOR-representations for the key perturbed quantities, such as the perturbed parallel plasma current density and the plasma radial displacement. SVD also helps to reveal the core versus edge plasma response to the applied 3D field. For the database covering the conventional aspect ratio devices, about 95% of data can be represented by the truncated SVD-series with inclusion of only the first five eigenstates, achieving a relative error (RE) below 20%. The MOR-data is further utilized to train neural networks (NNs) to enable fast reconstruction of perturbed 3D equilibria, based on the two-dimensional equilibrium input and the 3D source field. The best NN-training is achieved for the MOR-data obtained with a global SVD approach, where the full set of samples used for NN training and testing are stretched and form a large matrix which is then subject to SVD. The fully connected multi-layer perceptron, with one or two hidden layers, can be trained to predict the MOR-data with less than 10% RE. As a key insight, a better strategy is to train separate NNs for the plasma response fields with different toroidal mode numbers. It is also better to apply MOR and to subsequently train NNs separately for conventional and low aspect ratio devices, due to enhanced toroidal coupling of Fourier spectra in the plasma response in the latter case.

Study of spectral intensity of the laser ablated tungsten plasma and ablation mass at various laser spot sizes and laser fluence in vacuum environment

The ablation features such as spot size and crater volume are essential parameters for the in-situ laser-induced breakdown spectroscopy (LIBS) technique to be well applied in the quantitative diagnosis of the plasma-wall interaction (PWI) processes on the first wall or divertor in tokamak devices. The spectroscopic characteristics and the ablation features as well as the plume dynamics of laser-induced tungsten plasma in vacuum environment at various spot sizes were investigated in this work. A nano-second laser pulse (5 ns, 1064 nm) with a maximum energy of 200 mJ was focused on the tungsten target with different spot sizes (from 0.42 mm to 2.71 mm) by adjusting the distance from focal point to sample surface. A paraxially collective configuration (15°) and time-integrated spectral collection mode were applied, similar to the present in-situ LIBS system on Experimental Advanced Superconducting Tokamak (EAST). It was observed that the spot size affects the spectral intensity significantly, and the optimal spot size is 1.16 mm both for the observed tungsten ionic and atomic lines at the laser energy of 200 mJ. The electron temperature (Te), electron density (ne) and ablation mass also reach maxima at the optimal spot size. Temporally spectral dynamic investigation shows the tungsten line emission time at the optimal spot size lasts longer and decays more slowly than other sizes, and the fast photography investigation also verified this. It was found that the central intensity profile of the plasma plume image from the target surface along with the opposite direction of the laser incident at the early stage (< 8 ns, during the laser pulse) is stronger at the smaller spot size while it decays faster. This was attributed to the stronger plasma shielding effect leading to the relatively less total ablation mass. Moreover, the optimal spot size was found to increase with the laser energy in the energy range from 65 mJ to 200 mJ, and the corresponding optimal laser fluence almost keeps the constant value of ∼18 J/cm2, which provides valuable information for optimizing the current in-situ LIBS for quantitative diagnosis of the PWI processes such as fuel retention and impurity deposition in EAST.

Design of the Cooling Water System for the EAST Lower Divertor

The new lower tungsten divertor is a crucial component to achieve the targets of long pulse and high <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\beta $ </tex-math></inline-formula> H-mode operation of Experimental Advanced Superconducting Tokamak (EAST) device. The cooling water system (CWS) of lower divertor should have sufficient heat removal capability and be able to cope with pressure transients caused by the potential water hammer events, which makes the design challenging. In this study, the system process and details of main component are determined based on the design requirements. The thermal-hydraulic analysis model has been developed by Flowmaster (FLOWMASTER) to conduct the heat transfer analysis. Furthermore, the system water hammer transients under one pump trip condition have been analyzed. The results show that the flow distribution of each client is uniform, and the pressure and temperature of key points meet the design requirements. The maximum pressure fluctuation, 0.16 MPa, caused by the sudden closing of the check valve is within the acceptable range. The experiment data are in good agreement with the simulation values, which proves the correctness of the model and the feasibility of the design.

Determination of Eddy-Current Distribution in Electrically Isolated Vessel Sections of ADITYA-U Tokamak

The time-varying magnetic field induces eddy current in the conducting structure of a tokamak device. The characteristics of eddy currents in the discontinuous vacuum vessel are more complex than those found in a tokamak with a continuous vacuum vessel. Even with multiple electrically isolated sections of the vacuum vessel, there is evidence of a noticeable fast decaying current that flows along the inboard and outboard vessel surfaces and their analysis is not straightforward. It is necessary to detect and investigate the eddy-current effect due to the close vicinity of the vessel surface with the plasma column as they seriously affect the in-vessel magnetic diagnostics, the penetration of equilibrium fields, and the displacement of the plasma column. The induced eddy currents in vacuum-vessel sections are estimated using magnetic probes during the in situ calibration experiments in ADITYA-U tokamak. The filament model is utilized for estimating the spatial and temporal eddy-current profiles generated in the ADITYA-U vacuum vessel. The number of current filaments in the vessel section is optimized using the various combinations of available Mirnov sensors (magnetic probes) located over the poloidal periphery inside the vessel. In this article, the eddy-current measurement under different experimental conditions in the ADITYA-U is presented. This study also serves as an illustration of the usefulness of the filament model for the parameterization of eddy-current paths and magnitude to describe the induced vessel current during ADITYA-U plasma discharges.

Concept for a multi-purpose EU-DEMO pellet launching system

Pellets - mm-size solid bodies produced from frozen fuel – are mainly destined for fuelling purposes in the fusion reactor EU-DEMO. However, pellets have been proven capable tools for further tasks too, e.g. ELM frequency control and the efficient delivery of seeding gases have been already demonstrated. Here, a concept is presented for a single pellet launcher based on a stop-cylinder centrifuge accelerator equipped with multiple pellet sources. The sources can deliver pellets with different sizes and composition, finally combined into one single compound pellet train. Thus, the pellet launching system (PLS) is capable to control simultaneously different plasma parameters with a minimized cross talk between these different actuations. Currently, a new PLS is under development for the new large superconducting tokamak device JT-60SA which can be potentially regarded as a prototype for the envisaged EU-DEMO system. Its initial configuration will be capable to control plasma density and ELM pacing simultaneously; optionally a source for doped pellets can be added. Status and recent achievements of this systems are reported. At the full metal wall mid-size tokamak ASDEX Upgrade (AUG), work is ongoing developing a versatile control strategy and corresponding tools. Capable to inject pellets with high speed through a guiding tube from the torus inboard side, AUG represents a fully reactor relevant configuration. While all demonstrations have been performed with a single pellet source hence bound to actuation prioritizing one actuation parameter, all tools developed can be expanded straightforwardly to multi-purpose control. For example, one novel tool optimizes real-time feedback pellet flux control taking into account the discrete nature of the pellets. As well, a solution is worked to handle the issue of missed-out pellets. Such failed pellets are considered unavoidable due to the fragile nature of the solid fuel but identified as significant hazard for reactor burn control.

Divertor detachment of high β N H-mode discharges with the HL-2A closed divertor geometry

Divertor detachment of high H-mode discharges has been achieved by the injection of radiating impurities in the very closed divertor geometry of HL-2A. Non-intrinsic impurity is injected into the outer divertor chamber, and enhanced volumetric energy loss in the divertor and decreased divertor heat flux are then observed. Divertor detachment leads to a little degradation of the core confinement, characterized by slight decrease of plasma stored energy and pedestal pressure as well as increase of edge localized mode (ELM) frequency, but the high H-mode is still sustained with > 2 and H98 > 1. After impurity injection, the pedestal density fluctuation is increased, implying that the enhanced pedestal transport might be responsible for the degradation. During divertor detachment phase, the impurities are well controlled in the divertor without strong radiation losses near X-point region, and the main plasma density does not increase, but decreases slightly, which could be contributed by the very closed divertor geometry. These results are encouraging to achieve better heat load control and high plasma performance simultaneously for tokamak operations and to evaluate the effect of the divertor geometry for future tokamak devices.

Suppression of runaway current by electrode biasing and limiter biasing on J-TEXT

The avoidance of runaway electrons (REs) generated during plasma disruption is of great concern for the safe operation of tokamak devices. Experimental study on the suppression of runaway current by electrode biasing (EB) and limiter biasing (LB) has been performed on the J-TEXT tokamak, which could be an alternative way to suppress the runaway current. The experimental results show that the higher the voltage value, the smaller the runaway current in both EB and LB experiments. The runaway current can be completely suppressed at an electrode biased voltage of +450 V and a limiter biased voltage of +300 V. The comparison of the energy spectra during the runaway plateau phase shows that the maximum energy max (E RE) and radiation temperature T HXR of hard x-rays (HXRs) are significantly reduced after the application of +200 V limiter biased voltage. The electric field generated by the biased voltage may be the key factor to suppress the runaway current, and the measured radial electric field increases obviously after the voltage is applied. This may result in an increase in the loss of REs to realize the suppression of runaway current.

A new method to reduce the toroidal field ripple effect due to the ferromagnetic materials in updated CFETR

The toroidal field ripple (TFR) is one of the critical factors affecting the long-pulse high-performance discharges for the tokamak device. The TFR is not only determined by the toroidal field coils (TFCs) but is also affected by the ferromagnetic materials in the plasma facing components such as diagnostic system and blanket system. In the phase Ⅱ design of the China Fusion Engineering Test Reactor (CFETR), the inner ferromagnetic constituent and the poloidal segmentation of water-cooled ceramic breeder (WCCB) blanket are modified and the diagnostic system also has major change compared with the phase Ⅰ design, therefore it is necessary to investigate and reduce the TFR effect of the updated blanket system and diagnostic system for the phase Ⅱ designed CFETR. Firstly, the effect of the blanket system on the TFR is evaluated systematically, considering the diagnostic shielding modules (DSMs) of the diagnostic system at the equatorial ports (EPs). Moreover, a new TFR reducing method by adjusting the distribution of ferromagnetic materials filling ratios in the toroidal direction is proposed. According to the results, the local TFR nearby the plasma separatrix line already exceeds the design specification limit with the blanket system, and this phenomenon will be worse due to installing the DSMs and the EP. After adjusting the percentage of the ferromagnetic materials in the outboard-side (OBs) segment, the plasma TFR is reduced obviously and within the 0.5% limit.

Particle-in-cell simulations on parametric instability of the lower hybrid wave

Lower hybrid (LH) wave is an effective tool for current drive (CD) on tokamak devices. Parametric instability (PI) has always been a troubling phenomenon decreasing the lower hybrid current drive efficiency. In this work, the PI behavior of the LH waves for plasma parameters on the Experimental Advanced Superconducting Tokamak is investigated via first-principle simulation with a two-dimensional full-particle-in-cell method. The PI processes where an LH pump wave decays into another LH sideband and a low-frequency mode [ion sound quasi-mode (ISQM) or ion cyclotron quasi-mode (ICQM)] are observed in simulations. The ICQM-type and ISQM-type decay channels dominate, respectively, in the high- and low-plasma regimes while appearing simultaneously in the intermediate regime, which agrees well with the theoretical prediction by solving the parametric dispersion relation. In addition, for both the decay channels, their LH sidebands are excited with a wide range of parallel refractive index N∥ up to 30.0, which can resonantly interact with thermal electrons. The resulting electron heating can be enhanced due to overlap of resonances between the electrons and sidebands. Such electron heating not only dissipates the wave energy in the edge plasma but produces a significant portion of fast electrons, which may seriously affect tokamak discharges. The ion cyclotron heating due to the ICQMs is also observed in simulations.

Effect of relative locations between internal transport barrier and q min on HL-2M’s kink-ballooning stability

The significance of relative locations between the internal transport barrier (ITB) and the minimum value of safety factor (qmin) is proved by the magnetohydrodynamic (MHD) stability of ITB plasma, in shaped tokamak devices. In this work, equilibria of HL-2M tokamak with different locations of ITB and qmin are generated using the TOQ code, and the kink-ballooning stabilities of the equilibria with respect to 1 ≤ n ≤ 20 (n is the toroidal mode number) are simulated with the BOUT++ code. The results show that the MHD instability is suppressed magnificently, when the location of ITB is close to the core, while qmin is to the edge. The instability of the equilibrium is also stabilized when ITB is in the region with negative magnetic shear. As ITB moves toward the core or qmin moves toward the edge, the amplitude of negative magnetic shear increases, and the ITB gets closer to the negative magnetic shear, which benefits the MHD stability. Meanwhile, when qmin moves toward the edge, the elongation of the magnetic surface in the ITB region increases, and the area of the magnetic surface on the strong field side expands, which optimizes the magnetic surface distribution and improves the MHD stability.

Effect of continuously flowing liquid Li limiter on particle and heat fluxes during H-mode discharges in EAST

• Particle and heat fluxes were successfully controlled by using continuously flowing liquid Li (FLiLi) limiter in the H mode discharges with high plasma heating power in EAST. • By using flowing liquid Li limiter, the fuel recycling could effectively reduce by ∼ 50% and Zeff could decrease from 2.3 to 1.6. • The decreased recycling and impurity radiation led to the achievement of high energy confinement plasma with the average stored energy increased up to ∼ 290 kJ. • Due to Li vapor shielding effect, ∼40% plasma heat flux was dissipated before arrived at the Li limiter. Particle and heat fluxes were successfully controlled by using a continuously flowing liquid Li (FLiLi) limiter in the H-mode discharges with high plasma heating power in the Experimental Advanced Superconducting Tokamak device. There were strong interactions between the FLiLi limiter and high-power plasma with a ∼ 8.3 MW source heating power, and successively, a bright Li radiation ring was produced, which effectively decreased fuel particle recycling by approximately 50%. Due to Li efflux from FLiLi during a series of high-power discharges, an obvious real-time wall conditioning effect was produced, and fuel particle recycling further decreased. Moreover, the value of Z eff decreased from 2.3 to 1.6 due to a decrease in impurity sources; this was attributed to the accumulation of Li deposited on the first wall, which effectively protected the wall materials. The decreased recycling and impurity radiation achieved high-energy confinement plasma, and the average stored energy increased up to ∼ 290 kJ. Moreover, due to the effect of Li vapor shielding, nearly 30% plasma heat flux was dissipated before it arrived at the Li limiter. These results promote further exploration of liquid Li solutions for the critical challenge of heat flux handling and particle control in fusion power plants.

Py-EFIT: A new Python package for plasma equilibrium reconstruction on EAST tokamak

Plasma equilibrium is the basis for various tokamak research topics, such as plasma performance evaluation, plasma control, and physical analysis of instability, etc. In this study, we have developed a Python-based toolkit with a graphical user interface (GUI) to reconstruct the tokamak plasma equilibrium, which is called the Py-EFIT program. The main implementation steps of Py-EFIT program are introduced. Firstly, various magnetic diagnostic signals and machine structure of Experimental Advanced Superconducting Tokamak (EAST) are utilized as the inputs for computation. Secondly, the plasma magnetic flux distribution is numerically determined based on these inputs according to the plasma equilibrium reconstruction principle. Finally, a GUI of Py-EFIT program is developed for function execution and visualization of results. This program provides a convenient and user-friendly platform for plasma equilibrium calculation. It has been successfully applied for experimental analysis and plasma equilibrium reconstruction on EAST, which also indicates its potential for practical applications in more tokamak devices. Program summaryProgram title: Py-EFITCPC Library link to program files:https://doi.org/10.17632/85f8kj7hvp.1Licensing provisions: GNU General Public License 3Programming language: PythonNature of problem: Plasma equilibrium is the basis for many tokamak research topics, such as plasma performance evaluation, plasma control, physical analysis, etc. But plasma equilibrium cannot be obtained from experimental measurements directly, it has to be calculated through ‘equilibrium reconstruction’ technique. This program (Py-EFIT) provides an efficient, open-source, user-friendly, expandable, and portable computation platform for equilibrium reconstruction calculation.Solution method: 1. Code initialization, including grid mesh division and parameter declaration; 2. Give flux function an initial guess; 3. Construct plasma current matrix and response matrix; 4. Solve for plasma current based on diagnostic signals; 5. Calculate updated flux function based on Green-function method; 6. Refine flux value at magnetic axis and X-points; 7. Evaluate the difference between two iteration steps; 8. Output results until convergence condition is satisfied.

The Vlasov–Poisson–Landau System with the Specular-Reflection Boundary Condition

We consider the Vlasov–Poisson–Landau system, a classical model for a dilute collisional plasma interacting through Coulombic collisions and with its self-consistent electrostatic field. We establish global stability and well-posedness near the Maxwellian equilibrium state with decay in time and some regularity results for small initial perturbations, in any general bounded domain (including a torus as in a tokamak device), in the presence of specular reflection boundary condition. We provide a new improved \(L^{2}\rightarrow L^{\infty }\) framework: \(L^{2}\) energy estimate combines only with \(S_{ }^{p}\) estimate for the ultra-parabolic equation.

Study of the aluminum ablation features and spectral intensity at a various sample temperature in vacuum environment

For the LIBS in-situ wall-analysis in the tokamak devices, the detection is under a vacuum environment and the sample temperature changes in real time. The influence of the sample temperature (Ts) on the LIBS spectral intensity and ablation features could be paid much attention to ensure the LIBS analysis accuracy. In this study, an aluminum sample was chosen as a proxy of beryllium, which has been selected as the first wall in ITER device. A comparison was performed between the obtained emission in vacuum and in air whilst the sample was heated. The emission intensification as a function of Ts in vacuum is quite different from that in air at atmospheric pressure. To study the Ts on the intensification process in vacuum, plasma properties (temperature T, electron number density ne and plasma size) and laser ablation features (the ablated crater depth and volume) were investigated. The experimental results indicate that an increase in Ts leads to a significant enhancement in the spectral intensity of Al. The electron density ne of plasma increases with Ts, but the plasma temperature T almost does not change with Ts. Plasma plume images also remained nearly unchanged with Ts in the detecting time window. However, it was found that the laser ablated crater features (diameter, depth and volume) depend significantly on Ts. Increasing Ts results in a larger and deeper ablation crater and a higher ridge around the crater for the same laser fluence. Net ablated mass also increases with increasing Ts. The results should be helpful to improve the accuracy of the LIBS analysis for the first wall composition in the tokamaks such as EAST and ITER.

Optical filter-fiber combination based infrared measurement device for surface temperature measurement.

Neutral Beam Injection (NBI) is one of the four efficient ways of assistant heating on the Experimental Advanced Superconducting Tokamak (EAST) device. Measures have been taken to ensure more reliable monitoring of NBI components' thermal data. Infrared thermometers, unlike surface-mounted thermocouples, have the advantage of non-contact and thus will not cause interference and get far less chance of malfunctioning [Childs etal., Rev Sci. Instruments 71, 2959-2978 (2000)]. In this paper, the remote infrared temperature measurement with optical fiber is reported to meet the requirement of the EAST-NBI diagnostic system. Some tests have been conducted on the remote infrared temperature measurement under the working condition, and an atmospheric propagation simulation was performed to estimate the atmosphere absorption. The results indicate that the detector's response is in line with the theory, and the atmosphere absorption within a certain wavelength should not be ignored. Temperature measurement error is within one percent at temperatures above 200 °C. So, for the temperature range of 200-600 °C, the remote infrared temperature measurement by the optical fiber can fully meet its expectation.