• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3614-3619
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• 2022

To operate the ion cyclotron resonance heating (ICRH) antennas in a better heating state and produce relatively low impurities, it is necessary to control the antenna spectrum by changing the antenna phasing. As the electrical length of the antenna feeding transmission lines is changing as a matter of the standing wave pattern at the ceramic supports, 90° elbows, T-connectors and antenna loops, we chose to measure the current at the grounding points of the antenna loops by antenna strap probe. The voltage drops along a small, several millimeter-long paths at the end of the antenna loops give a signal that is proportional to the current in the antenna loop. Through the simulation of the antenna strap probe and the actual measurement of the antenna phasing under vacuum conditions, the reliability of the antenna strap probe based diagnostic system have been successfully proved. Moreover, this system was successfully applied to the ICRH daily experiments in the spring of 2021. In the near future, the active real-time feedback control of the antenna phasing system will be developed based on this diagnostic system in the EAST tokamak.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.207-219
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• 2022

The suppression of high-intensity blob structures in the scrape-off layer (SOL) by ion-cyclotron range of frequencies (ICRF) power, leading to a decrease in the turbulent fluctuation level, is observed first in the Experimental Advanced Superconducting Tokamak (EAST) experiment. This suppression effect from ICRF power injection is global in the whole SOL at EAST, i.e. blob structures both in the regions that are magnetically connected to the active ICRF launcher and in the regions that are not connected to the active ICRF launcher could be suppressed by ICRF power. However, more ICRF power is required to reach the full blob structure suppression effect in the regions that are magnetically unconnected to the active launcher than in the regions that are magnetically connected to the active launcher. Studies show that a possible reason for the blob suppression could be the enhanced E_r × B shear flow in the SOL, which is supported by the shaper radial gradient in the floating potential profiles sensed by the divertor probe arrays with increasing ICRF power. The local RF wave power unabsorbed by the core plasma is responsible for the modification of potential profiles in the SOL regions.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.216-221
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• 2024

Ion cyclotron range of frequency (ICRF) heating system is an important auxiliary heating method in the experimental Advanced Superconducting Tokamak (EAST). In EAST, several megawatts of power are transmitted with coaxial transmission lines and coupled to the plasma. For the long pulse and high power operation of the ICRF waves heating system, it is very important to effectively control the power and initial phase of the ICRF signals. In this paper, a power and phase feedback control system is described based on field programmable gate array (FPGA) devices, which can realize complicated algorithms with the advantages of fast running and high reliability. The transmitted power and antenna phase are measured by a power and phase detector and digitized. The power and phase feedback control algorithms is designed to achieve the target power and antenna phase. The power feedback control system was tested on a dummy load and during plasma experiments. Test results confirm that the feedback control system can precisely control ICRF power and antenna phase and is robust during plasma variations.

• Proceedings of the SAREK Conference
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• 2006.06a
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• pp.596-601
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• 2006

Because of insufficiency of energy resources and pollution of environment, it is necessary to develop alternative energy sources. Nuclear fission energy is used widely for source of electric Power but being restricted due to radioactivity problem. Nuclear fission is highlighted as the new generation of nuclear energy and researched worldwide because of low risk of radiation effect. The representatives of fusion research is China's EAST, KSTAR of Korea and ITER of world. Korea Superconducting Tokamak Advanced Research(KSTAR) project is on progress for the completion in August, 2007. In this study, the research of utility system for KSTAR be carried out. The utility system of KSTAR is consist of water cooling & heating system, $N_2$ gas system, DI water system, service water system and instrument air & auto control system. The progress of KSTAR utility system is under commissioning state after construction completion. The optimal operation scenario will be verified during commissioning and adopted to the KSTAR operation.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2621-2627
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• 2023

In Experimental Advanced Superconducting Tokamak (EAST) experiments, to achieve long pulse and high-power ICRH system operation, a new kind of ICRH antenna has been designed. One of the most critical factors in limiting the operation of long pulse and high power is the intense heat load in the front face of the ICRH antenna, especially the Faraday Screen (FS). Therefore, the cooling channels of FS need to be designed. According to thermal-hydraulic analysis, the FS tubes are divided into several groups to achieve more excellent water cooling capability. The number of series and parallel tubes in one group is chosen as six. This antenna went into service in the spring of 2021, and it is delightful that the temperature distribution of the FS tube is below 400 ℃ in 14.5 s and 1.8 MW ICRH system operation. However, the active water-cooling design was not carried out on the upper and lower plates of FS, which led to severe ablations on that region under long pulse and high power operation, and the temperature is up to 800. Therefore, the upper and lower side plates of the FS were designed with water cooling based on thermal-hydraulic analysis. During the 2022 winter experiments, the temperature of ICRH antenna FS was lower than 400 in the pulse of 200s and the power of 1 MW operation.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1501-1512
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• 2024

Plasma disruption in tokamak experiments is a challenging issue that causes damage to the device. Reliable prediction methods are needed, but the lack of full understanding of plasma disruption limits the effectiveness of physics-driven methods. Data-driven methods based on supervised learning are commonly used, and they rely on labelled training data. However, manual labelling of disruption precursors is a time-consuming and challenging task, as some precursors are difficult to accurately identify. The mainstream labelling methods assume that the precursor onset occurs at a fixed time before disruption, which leads to mislabeled samples and suboptimal prediction performance. In this paper, we present disruption prediction methods based on anomaly detection to address these issues, demonstrating good prediction performance on J-TEXT and EAST. By evaluating precursor onset times using different anomaly detection algorithms, it is found that labelling methods can be improved since the onset times of different shots are not necessarily the same. The study optimizes precursor labelling using the onset times inferred by the anomaly detection predictor and test the optimized labels on supervised learning disruption predictors. The results on J-TEXT and EAST show that the models trained on the optimized labels outperform those trained on fixed onset time labels.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4134-4145
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• 2022

The collimator is one of the high-heat-flux components used to avoid a series of vacuum and thermal problems. In this paper, the heat load distribution throughout the collimator is first calculated through experimental data, and a transient thermodynamic simulation analysis of the original model is carried out. The error of the pipe outlet temperature between the simulated and experimental values is 1.632%, indicating that the simulation result is reliable. Second, the model is optimized to improve the heat transfer performance of the collimator, including the contact mode between the pipe and the flange, the pipe material and the addition of a twisted tape in the pipe. It is concluded that the convective heat transfer coefficient of the optimized model is increased by 15.381% and the maximum wall temperature is reduced by 16.415%; thus, the heat transfer capacity of the optimized model is effectively improved. Third, to adapt the long-pulse steady-state operation of the experimental advanced superconducting Tokamak (EAST) in the future, steady-state simulations of the original and optimized collimators are carried out. The results show that the maximum temperature of the optimized model is reduced by 37.864% compared with that of the original model. The optimized model was changed as little as possible to obtain a better heat exchange structure on the premise of ensuring the consumption of the same mass flow rate of water so that the collimator can adapt to operational environments with higher heat fluxes and long pulses in the future. These research methods also provide a reference for the future design of components under high-energy and long-pulse operational conditions.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4102-4110
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• 2022

Aiming at high-power and long-pulse operation up to 1000 s, some improvements have been made for both 2.45 GHz and 4.6 GHz lower hybrid (LH) systems during the recent 5 years. At first, the guard limiters of the LH antennas with graphite tiles were upgraded to tungsten, the most promising material for plasma facing components in nuclear fusion devices. These new guard limiters can operate at a peak power density of 12.9 MW/m². Strong hot spots were usually observed on the old graphite limiters when 4.6 GHz system operated with power >2.0 MW [B. N. Wan et al., Nucl. Fusion 57 (2017) 102019], leading to a reduction of the maximum power capability. With the new limiters, 4.6 GHz LH system, the main current drive (CD) and electron heating tool for EAST, can be operated with power >2.5 MW routinely. Long-pulse operation up to 100 s with 4.6 GHz LH power of 2.4 MW was achieved in 2021 and the maximal temperature on the guard limiters measured by an infrared (IR) camera was about 540 ℃, much below the permissible value of tungsten material (~1200 ℃). A discharge with a duration of 1056 s was achieved and the 4.6 GHz LH energy injected into the plasma was up to 1.05 GJ. Secondly, the fully-active-multijunction (FAM) launcher of 2.45 GHz system was upgraded to a passive-active-multijunction (PAM), for which the density of optimum coupling was relatively low (below the cut-off value). Good coupling with reflection coefficient ~3% has been achieved with plasma-antenna distance up to 11 cm for the new PAM. Finally, in order to eliminate the effect of ion cyclotron range of frequencies (ICRF) wave on 4.6 GHz LH wave coupling, the location of the ICRF launcher was changed to a port that is located 157.5° toroidally from the 4.6 GHz LH system and is not magnetically connected.