HL-2M tokamak is a new-generation magnetic confinement fusion plasma device in China, which has realized the high parameter operation mode with 1 MA plasma current.

This study aims to investigate the turbulent transport associated with the internal transport barrier (ITB) using gyrokinetic calculations.

Numerical simulation was performed based on the gyrokinetic theory. The turbulent transport relevant to ITB was studied using the gyrokinetic toroidal code (GTC) combined with the equilibrium of HL-2M tokamak. A filtering zonal flow was considered in analyzing the influence of zonal flow on turbulent saturation level. A time evolution analysis of turbulent poloidal spectrum was conducted to investigate the effect of different wavelength modes on turbulent transport.

The results show that the turbulent transport at ITB saturates twice in succession, and the calculated average ion heat transport diffusivity is approximately twice that of the first saturation level. Moreover, the short-wave mode k_θρ_i~2.15 dominates the first turbulent transport saturation, whereas the long-wave mode k_θρ_i~0.49 dominates the second turbulent transport saturation. Specifically, an "M" shape distribution of the radial heat transport diffusivities is obtained at the transport barrier position during the turbulent saturation period. Finally, the minimum radial heat transport diffusivity during the turbulent saturation period occurs near the ITB where a maximum plasma temperature and density gradient occurs.

The turbulent transport at ITB may be dominated by two types of microinstabilities at different stages of turbulence development. Turbulent energy of the system is inversely cascaded from the modes with short wavelengths to those with long-wavelengths. The results are in good agreement with the theoretical prediction of the ITB.

In a tokamak, when fast electrons are deconfined by the tokamak magnetic field constraint and lost to the vacuum wall or limiter, the device may become damaged and the discharge may be affected.

This study aims to explore the loss behavior of fast electrons during discharge using a diagnostic system based on a ZnS(Ag) scintillator probe for detecting the loss of fast electrons on the Experimental Advanced Superconducting Tokamak (EAST).

The Geant4 simulation program was employed to simulate the interaction between electrons in different initial states and the scintillator probe of the diagnostic system. Firstly, the probe model and the filling material model of stainless steel and ZnS(Ag) coating were established in Geant4. Then, the interaction between electron beam and scintillator probe under different incident conditions (incident energy, angle, scintillator thickness, magnetic field size, etc.) were simulated, and the physical processes were recorded. Finally, the recorded data were accessed by MATLAB programming for analysis.

The results show that the contribution of secondary electrons and initial electrons to the luminescence intensity of scintillators occupies different dominant energy ranges. The luminescence intensity first increases and then decreases with the increase of incident electron energy, with a peak value around 12 MeV, and the number of emitted photons at oblique incidence is greater than that at vertical incidence. When the electron energy is lower than 4.3 MeV, secondary particles dominate the scintillation, and when the electron energy is higher than 4.3 MeV, primary particles dominate. The thickness of the scintillator has no significant effect on the peak position. After, the luminous intensity is considerably affected by the magnetic field angle and electron pitch angle after adding a magnetic field.

The results of this study contribute to the understanding of the fast electron loss signal detected by the scintillator probe in the EAST experiments, providing a basis for further study of fast electron loss.

Negative ion sources driven by radio frequency (RF) waves have become the preferred solution for future neutral beam injection systems.

This study aims to monitor the plasma discharge state of each exciter of a high power RF negative ion source by developing a plasma luminescence monitoring system based on a photodiode is designed and constructed.

The intensity of plasma emission was closely related to the number of specific collisions, collision particle density, and collision particle energy. Therefore, the intensity of plasma emission was applied to monitoring plasma parameters qualitatively, and a photodiode-based multichannel plasma luminescence monitoring system was designed and implemented to monitor the plasma discharge state of each exciter of a high power RF negative ion source. Based on a reasonable collision radiation model, plasma parameters were quantitatively obtained by analyzing the intensity of plasma characteristic spectral lines, and the influence of the filtering magnetic field generated by plasma current on plasma emission signals as experimental tested.

Experimental results this monitoring system demonstrate that the real-time intensity information of plasma emission from different positions is successfully collected, and subsequently presented and saved in the form of voltage signals for real-time monitoring and post-data processing by the host computer. The intensity of plasma emission has a good linearity with RF discharge power.

The plasma light monitoring system of this study can accurately and real-time measure the excitation, maintenance, and extinction processes of plasma in the RF exciter.

The control of edge localized modes (ELMs) is a key issue for the safety operation of ITER and future magnetic confinement fusion reactors. In recent years, extensive theoretical simulations and experimental studies have demonstrated that resonant magnetic perturbation (RMP) is a promising method for controlling edge localized modes (ELMs) in H-mode plasmas.

This study aims to investigate drift kinetic resonant effects of thermal particles on the plasma response to the applied RMP and compare with that of the fluid model for better understanding of ELMs control in HL-2A.

Based on experimental plasma and RMP coil configurations in HL-2A, the MARS-F/K codes were employed to compute the drift dynamic response of plasma to RMP under high constraint mode and compared it with the results of fluid model. Further sensitivity studies were conducted on key parameters including the plasma equilibrium pressure, toroidal flow as well as the thermal particle collision effects.

The fluid response model predicts an initially relatively weak enhancement of the plasma response amplitude with the equilibrium pressure parameter βN, followed by a strong variation as βN approaches the Troyon no-wall beta limit. The latter is unphysical and is eliminated by the kinetic response model. Including kinetic response and considering particle collision, the non-adiabatic resonant contribution of trapped thermal ions is found to be significant. Neglecting the particle collision, however, the trapped thermal electron contribution becomes more pronounced.

The importance of considering kinetic effects for high-beta plasma response is emphasized by this study. It is important to include the non-adiabatic resonant contribution of trapped particles to the kinetic response while taking into account the particle collision effect.

When fusion nuclear reactor is in operation, neutron activation causes a large number of radioisotopes. Activation calculation is a very important step in both of reactor shielding calculation and radiation safety analysis.

This study aims to develop the capability of transport-activation coupling calculation for fusion reactor based on the Monte Carlo code cosRMC.

Firstly, the built-in burnup solver "Depth" of cosRMC was employed to develop transport-activation internal coupling calculation function under fixed source mode with embedded calculation of activation related nuclide single group reaction cross-sections in neutron transport process without the transmission of neutron spectra to external activation programs. Then, the developed code was applied to the activation calculations of the first wall (FW) material steel and plasma facing component (PFC) material tungsten of the Chinese Fusion Engineering Testing Reactor (CFETR) using continuous energy cross-sections and multi group cross-sections, respectively, and calculated results were compared and verified with that of the activation program ALARA.

The comparison results of activation calculation of FW steel and PFC tungsten in CFETR show the consistency between cosRMC-based internal coupling method and ALARA program, which preliminarily verifies the correctness of the transport activation internal coupling calculation function of the developed cosRMC program.

The developed cosRMC-based internal coupling method can dynamically update neutron spectra and material information, and use continuous energy cross-sections for reaction rate calculation, obtaining reaction cross-sections related to the geometry and energy spectra of actual problems, thus accurately considering the influence of resonance zone nuclear cross-sections.

With the development of human aerospace industry, it is necessary to develop propulsion systems suitable for different space mission scenarios. MegnetoPlasmaDynamic thruster (MPDT), which is similar to the principle of magnetic confinement fusion, is a typical representative of electromagnetic thruster, which stands out among many electric thrusters because of its superior performance in thrust power ratio and specific impulse. Anode power deposition is the result of the interaction between plasma and wall during MPDT operation. It is one of the main mechanisms of power loss of this type of thruster, accounting for 40%~90% of the total power, which seriously reduces the efficiency of thruster.

This study aims to solve the problem of low efficiency of thruster by investigating the influence of anode radius on the efficiency of thruster from the perspective of anode power deposition.

First of all, based on MagnetoHydroDynamic (MHD) equations, numerical models for radial discharge parameters and physical model for anode power deposition were established. Then, the influences of anode radius on discharge parameters, anode power deposition and anode power deposition fraction were studied on the basis of these models by numerical calculation method. Finally, a water-cooled structure anode was designed, and the effectiveness of its heat dissipation structure was verified by thermal simulation.

The results show that with the increase of anode radius, the electron density and ion velocity are increased, the anode power deposition fraction is decreased whilst the anode power deposition is increased. The thruster efficiency is improved by increasing the anode radius. The thermal simulation results show that when the input power deposition of the water-cooled structure anode is about 3 kW, the corresponding temperature difference of the anode cooling water is 5 K.

This study verifies the reliability of the physical model of anode power deposition, indicating that increasing the anode radius is an effective means to improve the efficiency of the thruster.

Electric propulsion systems, compared to traditional chemical propulsion, offer longer operational lifespans and lower fuel consumption in space missions, garnering significant attention in recent years. However, for high-power Hall effect electric thrusters developed based on controlled fusion concepts, measuring thrust proves challenging due to the high-temperature environments required for ionizing propellants in electric thruster, resulting in the generation of hot plasma plumes during operation. As a result, traditional thrust measurement methods are unable to accurately measure the thrust.

This study aims to accurately measure thrust during ground testing of electric thrusters for precise control of the spacecraft's attitude and orbit maintenance.

Firstly, a thrust measurement platform based on flexible beam structure was designed and implemented with capability of measuring the thrust generated by electric thrusters in high-temperature. Simultaneously, a calibration system was built to verify the stability and repeatability of the thrust measurement results. Then, simulation analysis was conducted on the structural mechanics and thermal coupling field of the thrust measurement system under different operating conditions. Finally, the variable specific impulse magnetoplasma rocket (VASIMR) was taken for experimental thrust measurement of its electric propulsion system.

The experimental results of VASIMR indicate the thrust is 266.5 mN measured in real time at the central magnetic field intensity of 0.2 T with a mass flow rate of 20 mg?s^-1.

The thrust measurement platform based on the bending beam structure can meet the measurement requirements of VASIMR, providing valuable references for subsequent experiments.

Neutral beam injection heating is a very effective auxiliary heating method in magnetic confinement fusion experiment. The built 5 MW neutral beam injection plays an important role in the H-mode operation with 1 MA plasma current of HL-3 tokamak. In order to achieve the H-mode operation with 2.5 MA, three neutral beam injection heating systems will be constructed with a total heating power of 20 MW.

This study aims to investigate the transmission efficiency of 7 MW neutral beam injector designed for HL-3 tokamak.

The beam transmission efficiency was studied from two aspects of neutralization efficiency and power deposition. The initial target thickness of the neutralizer under two beam parameters was analyzed based on Monte Carlo method. The optimal neutralization efficiency was achieved by adding neutralizer gas supply, and the target thickness required for the optimal neutralization efficiency was given. Finally, power deposition method was employed to analyze the transmission performance of the two neutral beams with different parameters at different beam divergence angles.

The results show that the optimal neutralizer efficiency is achieved when the neutralizer gas supply is 0.6 Pa?m³?s^-1 and 1.1 Pa?m³?s^-1, respectively. The neutral beam power obtained under two beam parameters with a divergence angle of 1.2° is 7.2 MW and 6.8 MW, respectively, and the corresponding transmission efficiency is 0.4 and 0.35, respectively.

The new neutral beam injector of HL-3 tokamak will be built according to the design of this study, with optimized parameter of 100 kV/45 A, while still retaining the ability to deliver a 120 keV deuterium beam.

In neutral beam injectors (NBIs) of Tokamak device, the calorimetric target is one of the most important water-cooled components, responsible for receiving and measuring beam power. In addition, by using a built-in thermocouple array, the temperature rise at different positions of the target plate can be monitored in real-time, thereby obtaining the power density distribution of the extracted ion beam or neutral beam.

This study aims to develop a calorimetric target for the neutral beam injector in the HL-3 device, which can meet the requirements of target plate lifting and thermal load absorption.

A linear push rod mechanism was applied to the design of the calorimetry target to achieve lifting and lowering, and a "W" - shaped target plate structure was adopted to achieve absorption of neutral beam energy. According to the design parameters, the fluid calculation module of Ansys Workbench was employed to simulate the temperature distribution of the calorimetry target under full power operation.

Calculation results show that the maximum temperature rise of the calorimetry target is 526.4 oC when the deflecting magnet is opened under full power operating conditions, and the temperature can be lowered to room temperature within 30 s, satisfying the requirements for the use of the beam line.

The successfully developed calorimetric target of this study meets application requirements on second neutral beam injection beamline of HL-3 device, providing references for further engineering design of calorimetric targets in other NBIs.

ENN Science and Technology Development Co., Ltd. (ENN Fusion Technology R&D Center) is upgrading its compact fusion research facility EXL-50 to EXL-50U. Both devices are the conventional conductor tokamak, on which the magnet power supply system is composed by 1 TF (Toroidal Field) power supply, 1 CS (Center Solenoid) power supply and 10 PF (Poloidal Field) power supplies PF1-10. All 12 sets of power supply system are powered by 2 AC pulse generators and output DC current through thyristor-based converters.

This study aims to design EXL-50U magnet power supply for satisfying high parameter requirements of EXL-50U.

Power supply capacity was the first concern for upgrading and the corresponding protection strategies under high parameter conditions was taken into account as well. The configuration of AC pulse generator was introduced at the beginning. Then transformers and converters were listed and designed in scheme. Control system and protection process were implemented respectively, followed by detail power supply system illustration and commissioning waveforms display for each power supply.

The reliability and controllability of developed power supply system are verified by the waveforms that forms plasma current under the condition of CS breakdown.

It is proved that this power supply system can work stably, and output waveforms can be repeated no matter it works alone or under complex condition of joint debugging.

Due to its pollution-free nature and non consumption of fossil fuels, nuclear fusion is the most ideal future energy source. China is preparing to build a China Fusion Engineering Test Reactor (CFETR) with independent intellectual property rights, and plans to build a commercial thermonuclear fusion power plant that can generate electricity externally by the mid-20th century. However, there is contradiction between the instability of nuclear fusion heating power output and the smooth operation of steam turbine generators, hence thermal storage technology is used for peak shaving and valley smoothing of power output in nuclear fusion reactors.

This study aims to compare heat storage technologies applied to CFETR nuclear fusion power plants to reduce its the peak and valley power output.

The parameters of helium cooled ceramic breeder cladding in nuclear fusion reactors was selected as the boundary conditions for thermal storage technology. By evaluated the applicable temperature range of thermal storage technology, three potential thermal storage technologies, i.e., chemical heat storage, sensible heat storage technology and phase change heat storage, for CFETR nuclear fusion power plants were preliminarily analyzed, and their costs were preliminarily predicted.

The three major types of heat storage technologies can all select heat storage media suitable for the temperature parameters of the helium cooled breeder blanket in CFETR nuclear fusion power plants. However, chemical heat storage has the potential to be applied in CFETR nuclear fusion power plants due to the temperature difference between its heat absorption and release, which is not conducive to the stability of the system and causes energy loss. Sensible heat storage technology and phase change heat storage technology have smaller temperature differences between their heat absorption and release. The preliminary economic analysis results show that the cost of phase change heat storage is the lowest, followed by molten salt heat storage, and the use of silicon bricks as the heat storage medium in solid-phase sensible heat storage technology. The use of cast steel as the heat storage medium in solid-phase sensible heat storage technology has the highest cost.

In thermal storage technology, molten salt thermal storage technology has a high degree of maturity and has a large number of engineering applications, with a cost between phase change thermal storage and solid-phase sensible thermal storage, and has great potential for application. The cost of phase change heat storage is the lowest, and the parameters are suitable for nuclear fusion power generation. However, its technological maturity is relatively low, and it is expected to become a focus of future research.

Data-driven methods for power fault diagnosis heavily rely on the signal data quality of power sensors. The power systems in Tokamak fusion devices often operate in environments with complex electromagnetic field coupling, leading to the mixing of physical characteristic signals with a significant amount of inseparable noise in the collected data.

This study aims to mitigate the impact of noise on the final diagnostic results by proposing a multi-branch denoising network, termed Hierarchy Branch Denoising Convolutional Neural Network (HBD-CNN) that utilizes noise-resistant wavelet enhancement in conjunction with one-dimensional convolutional neural networks to accomplish power system fault diagnosis tasks under the influence of noise interference.

Firstly, the signal decomposition function of discrete wavelet transform (DWT) was incorporated into the network layer of the convolutional neural network (CNN), and the optimization of the traditional 1D-CNN network structure was deepened alongside the more robust exponentially linear unit (ELU) for noise. Then, a data multi-level structure was constructed based on prior knowledge to leverage and couple it with hierarchical classification modules within the network, hence the generalization capability of HBD-CNN was enhanced. Finally, preliminary validation of the architecture of this model was conducted based on the simulated power supply dataset.

Validation results show that the fault diagnosis accuracy for the power converter reaches 98.31% when the signal-to-noise ratio (SNR) is 10 dB. Even at an SNR of 2 dB, the accuracy remains above 92%.

The results of this study indicate that HBD-CNN demonstrates excellent fault diagnosis performance and potential under noisy conditions.

Lower hybrid current drive (LHCD) is one of the main auxiliary heating and current driving methods for Tokamak, and studies have shown that the boundary parasitics effect of lower hybrid (LH) wave in scrap-off layer (SOL) significantly decreases current drive efficiency of LH wave. Among these, wave scattering caused by density fluctuation at the boundary results in spectral changes within the SOL, which changes the power deposition location and current drive efficiency.

This study aims to explore the lower hybrid scattering caused by density fluctuation in the scrap-off layer on the Experimental Advanced Superconducting Tokamak (EAST) device.

Based on the cold plasma dispersion theory and the blob density fluctuation theory, COMSOL Multiphysics was employed to establish a full-wave solution model of lower hybrid scattering in the SOL. Based on the parameters of the EAST device, the influence of low-frequency electron density fluctuation with different characteristics on wave scattering was analyzed.

The simulation results show that the backscattering direction caused by the density wave packet "blob" is more obvious than the forward scattering, and scattering induced by blob leads to the change of spatial structure of low hybrid power flow. The density fluctuation in the blob mainly affects the amplitude of the wave field disturbance, the radius of the blob mainly affects the spatial range of the wave scattering, the full-field disturbance caused by multiple blobs increases significantly.

The full-wave solution model of the SOL established in this study is reliable. The simulation results using this model show that the density fluctuation will cause lower hybrid scattering, which will change the spatial structure of the power flow.

The high voltage power supply is an important part of the neutral beam injection heating system, which determines the beam energy and the quality of the extraction beam current. With the gradual increase of voltage level, the pulse step modulation (PSM) high voltage power supply cannot meet the experimental requirements.

This study aims to design an inverter high voltage power supply based on super capacitor energy storage to achieve fast switching of injected power for neutral beam modulation.

Super capacitor energy storage was adopted to reduce the required grid capacity and minimize the impact on the grid. The DC-DC resonant converter structure with soft-switching technology was used to improve the response speed of the power supply and reduce the switching loss of the switching devices. After the design of power module circuit topology,system modeling and calculation based on power supply performance specifications, the charging circuit and main loop PSIM simulation model were established, and the power supply performance indexes were simulated and verified. Finally, a test prototype of inverter power supply module was built to conduct the test of relevant performance indexes. [Results and Conclusions] After simulation and experimental verification, the power module is able to achieve a stable output of 1 600 V/50 A, which meets the design requirements of 6 MW/120 kV.

Electron cyclotron resonance heating (ECRH) is an important heating and plasma current control method for the HL-3 tokamak. Microwave inject into plasma through the launcher, which is an important part of the ECRH system. The ECRH launcher system of HL-3 tokamak consists of three launchers, including a mid equatorial launcher and two upper launchers. All launchers are located in the same sector, and can work together to complete functions such as heating and Neoclassical Tearing Mode (NTM) control.

This study aims to design and test the ECRH launcher system of the HL-3 tokamak.

First of all, the overall planning was carried out for the launcher system. Subsequently, the transmission path and structure of the launcher were designed, and the effect of microwave injection were calculated by simulation. Then, detail design and implementation for the mid-equatorial launcher and No.1 upper launcher were completed, so did the optical path design for No.2 upper launcher. Finally, the transmission angle and rotation speed of the launcher were tested, and the rotation angle of the launcher was calibrated.

The full range response time of the mid-equatorial launcher is less than 90 ms. The full range response time of the No.1 upper launcher is less than 190 ms.

The control of the equatorial launcher and the No.1 upper launcher is precise and fast, meeting the requirements for experimental use of the tokamak.