China Fusion Engineering Test Reactor Research Articles

Progress of Engineering Design of CFETR Machine Assembly

The major radius of the China Fusion Engineering Test Reactor (CFETR) is 7.2 m, and its minor radius is 2.2 m, which are larger than those of the International Thermonuclear Experimental Reactor (ITER). That makes the assembly of the CFETR machine more intricate and challenging due to the assembly tool design, and their stresses are more complex when the weight of key parts/components increases, especially the assembly of the cryostat vessel, the vacuum vessel (VV), the toroidal field (TF) magnets, the poloidal field (PF) magnets, and the thermal shielding (TS). Based on the characteristics of the CFETR machine, a 45-deg assembly sector (with eight sectors in total), containing one 45-deg VV sector, two TF magnets, and one 45-deg VV TS, was designed as an assembly unit together with its assembly tooling. To improve the assembly efficiency, three working regions along the toroidal direction of the CFETR machine were designed to operate simultaneously. In addition, the assembly tools of the PF magnets and the cryostat were considered, and all of them are capable of supporting and adjusting the large CFETR machine components. Meanwhile, to improve their assembly accuracy and measurement efficiency, a laser tracker, an indoor global positioning system, and a scanner were employed in their assembly process. In addition, a metrology network was built for assembly of the CFETR machine.

Events detection and handling based on hierarchical finite state machines in EAST PCS

Experimental Advanced Superconducting Tokamak (EAST) is committed to long pulse high performance operation where plasma parameters will necessarily approach stability limits, which sets a higher demand on real-time events detection and handling (EDH). In order to enhance the current EAST plasma control system (PCS) events handling ability and explore a new intelligent scheduling mode for China Fusion Engineering Testing Reactor (CFTER) PCS, the EDH algorithm based on hierarchical finite state machines (HFSM) is designed in EAST plasma control system (PCS). The events and states in PCS were designed, and we successfully validated the state transition and algorithm scheduling. Moreover, a new event detection and processing for controlling hot spots of in-vessel components from hot spots is developed, and a simulation experiment is carried out in the closed-loop test environment between PCS and MATLAB/Simulink. The simulation results show that the EDH algorithm could capture the exception in time and implement effective strategies with the change of hot spots temperature. The temperature of hot spots were prevented from reaching a drastic situation, which proved the effectiveness of the EDH to protect the internal components of the device.

Investigation and Analysis on Tritium Accumulation Phenomenon Based on a Water-Cooled Blanket for CFETR at 1.5 GW

China Fusion Engineering Test Reactor (CFETR) is a major scientific project independently designed and developed by China, which will demonstrate the engineering feasibility of continuous large-scale fusion energy for safe and stable power generation. One of the most critical issues for fusion reactors is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.0, as it relates to the ability of the fusion reactor to fulfill the fuel cycle and continue stable operation. Moreover, the release and inventory of tritium is also a significant factor affecting the operation of CFETR. The temperature distribution inside the breeder blanket is precisely related to the tritium release. Therefore, it is particularly important for CFETR to optimize the temperature field inside the breeding blanket. It is significant that to ensure that the temperature of the mixed pebble bed zone reaches the limit of 900 °C as much as possible, which could maximize the tritium release efficiency. Simultaneously, it must make sure that other structural materials meet the allowable value of temperature. Tritium accumulation close to the cooling pipe is investigation, which will be adverse to tritium release. Hence, in order to solve this difficulty, the Be12 Ti tube was sleeved on the cooling pipe to increase the temperature of the mixed pebble bed zone near the cooling pipe. In this work, a water-cooled solid breeder blanket was proposed and the nuclear heat was calculated based on the 1.5 GW fusion power. The nuclear heat was used as the preprocessing data for thermal calculation carried out by the Monte Carlo N-Particle Transport Code (MCNP). With the given conditions, the thermal hydraulic was simulated with the commercial CFD code, the ANSYS CFX. In this way, a breeding blanket structure with a uniform temperature field of mixed bed zone close to 900 °C can be obtained. In addition, the selection of Be12 Ti tubes with the thickness of 0.525 cm can ensure a high TBR and increase the temperature around the cooling tubes to improve the tritium release efficiency. This work provides an important reference for the subsequent engineering design of CFETR breeding blanket.

Numerical study on influences of parallel sonic wave damping, precession, and bounce resonances on resistive wall modes in CFETR

Adopting an 11 MA steady-state reversed magnetic shear equilibrium of the China Fusion Engineering Test Reactor, the influence of parallel sonic wave damping, precession, and bounce resonances on the resistive wall mode (RWM) at different plasma pressures and plasma toroidal rotations is numerically investigated using the MARS-K code. We find that the parallel sonic wave damping, precession, and bounce resonances have damping effects on the RWM. The RWM is stabilized by the parallel sonic wave damping at low plasma pressure. At the two q = 3 rational surfaces, the RWM is the most unstable, but the precession resonance mainly occurs at one of the q = 3 rational surfaces, so the RWM cannot be stabilized. In addition, the precession resonance and bounce resonance occur simultaneously, and there are two roots. One is the precession resonance root, and another is the bounce resonance root. The bounce resonance has a stabilizing effect on the RWM.

Performance of highly flexible sub-cable for REBCO Cable-In-Conduit conductor at 5.8 T applied field

Due to the high current capability and excellent flexibility, High Flexible REBCO Cables (HFRC) have emerged as an important candidate for composite high-temperature superconducting conductors. The REBCO six around one Cable-In-Conduit Conductor (CICC) concept has been designed for application in the Central Solenoid (CS) coil of the China Fusion Engineering Test Reactor. In the application of fusion devices, the performance of CICC under electromagnetic (EM) loading and thermal stress is very important for reliable and economic operation. Therefore, a 1.22 m long sub-cable with HFRC design for CICC was manufactured and tested at 4.2 K in a background magnetic field up to 5.8 T. The aim is to investigate the stability of the current-carrying properties of the HFRC cable under electromagnetic and thermal cyclic loading. The test results show that the critical current ( I c ) of the HFRC cable reached 17.3 kA in a background magnetic field of 5.8 T at 4.2 K. Furthermore, no performance degradation was observed after 24 cycles of 80 kN/m peak load with a background field of 5.8 T and 8 warm-up-cool-down cycles between 77 K and room temperature. The test results provide a good basis for the development of full-size conductors in future magnet applications.

Structural Design of DC Magnet for Super-X Test Facility

A large scale research and development project for China Fusion Engineering Test Reactor (CFETR) named the comprehensive research facility for fusion technology (CRAFT) started in 2019 and is scheduled for completion in 2025. A superconducting conductor experiment testing platform (Super-X for short) is a sub-task of CRAFT to evaluate the future superconducting part performance. The maximum testing magnetic field at bore center is 15 T, the dimensions of testing space are 100 × 160 × 550 mm, the field homogeneity of testing space is superior to 95%, and the maximum testing current is 100 kA. The direct current (DC) magnet is the key sub-system for the facility providing the background magnetic field. The engineering design for DC magnet has been accomplished in the middle of this year, and its technical parameters meet the required. The main work focuses on the structural design of the split coils, the axial preload and the hanging assembly. This paper presents the progress on the engineering design of the DC magnet.

The Preload Analyses of CFETR CSMC Pre-compression Structure

CSMC is preparatory towards the final design and fabrication of China Fusion Engineering Test Reactor Central Solenoid. As one of the key components, the preload structure plays a critical function to maintain the integrity of CSMC winding modules under different load cases. Thus, it requires the preload structure has enough strength to resist the different kinds of external load. Moreover, to uniformly distribute the preload on the winding modules, it is needed to decrease the preload difference between the inner and outer tension rod under the steady operation state. In the paper, the preload variation under different load cases is analyzed and two different strategies are proposed to decrease the preload difference. The analysis results demonstrate the schemes are feasible. The methodology given in the paper is helpful for the design of the CFETR Central Solenoid.

Engineering design of the CFETR machine

The China Fusion Engineering Test Reactor (CFETR) is the next step in China's roadmap for magnetic confinement fusion development. It aims to bridge the gap between the ITER fusion experimental reactor and the demonstration fusion power reactor (DEMO). The CFETR will be operated in two phases. Phase I aims to achieve fusion power of 50–200 MW, Qplasma = 1–5, TBR > 1.0 and neutron irradiation effects ∼ 10 dpa. Phase II aims to perform tokamak DEMO validation under fusion power > 1 GW, neutron irradiation effects ∼ 50 dpa and figQplasma > 10. To ensure that the CFETR machine can achieve these goals, the major/minor radii of plasma have been altered from R = 5.7 m/a = 1.6 m to R = 7.2 m/a = 2.2 m. The new machine can meet the targets of both Phase I and Phase II and achieve smooth transfer between them. The configuration of the CFETR machine is presented in this paper. Then, the design progress and status of some key components are described, including the magnet system, blanket system, diverter system, remote handling system, vacuum vessel, thermal shield and cryostat. The research and development of most components are underway and will be demonstrated in the near future.

Development of temperature measurement control unit for cryosorption pump

For the future development needs of fusion, CFETR (China Fusion Engineering Test Reactor, referred to as CFETR) negative ion based neutral beam injector (referred to as NNBI) verification prototype is designed with large cryosorption pump as the main vacuum acquisition equipment. The temperature of the cryogenic adsorption surface and the radiation shield baffle need to be monitored in real time at key locations. In order to control the regeneration of the cryosorption pump, the ability to heat the cryogenic adsorption surface to a set temperature is required. A multi-channel temperature measurement control unit based on the STM32 was designed to meet the temperature measurement requirements of the cryosorption pump of the CFETR NNBI verification prototype, which operates in a complex radiation and electromagnetic environment. The temperature measurement range is from 4 K to 325 K in the helium zone and from 70 K to 450 K in the nitrogen zone. It not only measures 8 points simultaneously and transmits the data remotely to the upper computer, but also allows the temperature to be set by the upper computer. In addition, it is possible to control the regeneration process of the cryosorption pump by controlling the heating of the heating wire via the STM32. The temperature measurement control unit has been tested on a small cryosorption pump bench, it is capable of multi-channel temperature measurement and real-time display on the upper computer's display system,it can also control the heating. It not only lays the foundation for the construction of a temperature measurement control unit for the NNBI verification prototype cryosorption pump, but also provides a design reference for similar requirements.

Study on the Heat Absorption Coefficient of Tungsten as the Plasma-Facing Material for the Upgraded EAST Lower Divertor

Divertor is one of the key plasma-facing components (PFCs) for Experimental Advanced Superconducting Tokamak (EAST) since they play important roles in resisting the energetic particle irradiation and huge heat deposition from the core plasma as well as efficaciously dispelling such energy load. Tungsten is used as the plasma-facing material for PFCs due to the excellent properties. During the thermal analysis and the high heat flux test of the PFCs, the heat absorption coefficient is an important parameter that directly affects the availability and reliability of the design structure. Previous studies have shown that there is a large deviation in the measured heat absorption coefficient of tungsten. Therefore, it is very important to accurately measure the heat absorption coefficient of tungsten. In this article, the heat absorption coefficient of tungsten was tested by the e-beam device of the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). In this test, 36 sets of inlet and outlet water temperature data were obtained by changing the water flow rate and the incident heat flux. The heat absorption coefficient of tungsten was calculated by the theoretical calculation using water calorimetry. The results indicate that the heat absorption coefficient of tungsten is stable at ~0.54, which can be used as a reference for the engineering design of PFCs of the upgraded EAST lower divertor. All test parameters for EAST will provide important reference and guidance to the next-generation fusion facility China Fusion Engineering Test Reactor (CFETR).

Numerical investigation of ECCD under the CFETR concept design parameters

RF power in the electron cyclotron (EC) frequency range is planned to be injected to the plasmas of China Fusion Engineering Test Reactor (CFETR). Numerical investigation for electron cyclotron current drive (ECCD) is performed under the parameters of the latest/final version of the CFETR concept design (R = 7.2 m, a = 2.2 m, BT = 6.5 T). The TORAY-GA ray-tracing code is used to optimize frequency and launcher ports of EC waves. It is found that both 240 GHz and 250 GHz EC waves can meet the desired current drive demand in the target radial region (ρ ≤ 0.40). The 240 GHz EC wave is the optimum candidate. It can be launched either from equatorial launcher (EL) port, upper launcher (UL) port or top launcher (TL) port. EC waves with frequency equal to or larger than 250 GHz will be suffered strong parasitic second harmonic absorption if it is launched from EL port. Except for 250 GHz wave launched from EL port, ECCD efficiency increases with the increasing of EC frequency in the range of 190 ∼ 250 GHz. The maximum CD efficiency γCDn20RIec/Pin can reach 0.3 taking account for improvement of the CD efficiency about 20% enhanced by momentum conserving correction in the central radial region (ρ ≤ ∼ 0.20), either by 240 GHz EL-ECCD or 240/250 GHz UL-ECCD launching from (Ri, Zi) = (8.2, 4.5) m. Large toroidal incident angle α is needed for higher EC frequency to obtain an efficient CD efficiency. Necessary and sufficient conditions for successful and effective TL-ECCD are given. Compared to UL-ECCD, TL-ECCD does not significantly improve the efficiency of ECCD in the mid-radius region (ρ = 0.4 ∼ 0.6). Most cases of TL-ECCD significantly improve the efficiency of ECCD in the mid-radius region compared with EL-ECCD, but some cases does not. It is different from the results of previous theoretical simulations, in which TL-ECCD shows significantly improve ECCD efficiency than EL launcher. These findings are of significance for design of an efficient ECCD system under the CFETR design parameters.

Design and implementation of a dynamic system simulation framework for plasma control system verification platform

The plasma control system (PCS) is a key part in tokamak operations, which requires test and verification. Dynamic system simulation tools have been widely utilized to build a simulation environment to verify the design of plasma control systems as well as other parts of tokamak systems under various operation scenarios, save the time and cost of real discharges, improve the efficiency of deployment and reduce risks. To provide a simulation platform based on open-source software as part of the infrastructure of the PCS-VP (Verification Platform) of the China Fusion Engineering Test Reactor (CFETR), a framework named DSSim (Dynamic System Simulation) is proposed. It is an environment for dynamic system simulation based on Python, which enables users to build and simulate hierarchical block diagram models with interconnected systems that corresponds to different dynamic systems. Solutions for special tasks that helps improve the accuracy of simulation like flattening the model hierarchy, scheduling system execution order and solving continuous systems are implemented and will be automatically invoked in different phases of simulation. A GUI client with the DSSim framework built in is developed to allow for constructing and simulating block diagram models in a more user-friendly way. A model of the vacuum field system of the Experimental and Advanced Superconducting Tokamak (EAST) is built and simulated with DSSim and the result is compared with that of the same model in MATLAB/Simulink to test the framework's precision and usability.

Mechanical characterization and performance of reduced scale mock-up for the CFETR first wall manufactured by SLM process

The blanket, one of the most important parts in China Fusion Engineering Test Reactor (CFETR), is the carrier with core function for fusion reactor application. First wall (FW) is the key unit of the blanket, facing the plasma directly and playing a decisive role in the fusion reactor. In this study, the reduced scale FW U-shaped mock-up formed by selective laser melting (SLM) was investigated. The influence of microstructure anisotropy on mechanical properties from vertical direction (z-x) and horizontal direction (x-y) was researched and the feasibility of SLM technology applied to nuclear fusion was explored. Experimental results show that the grain structure is mainly polygonal cellular crystals in x-y direction and elongated columnar crystals in z-x direction under scanning electron microscope (SEM) observation. This grain growth mode is attributed to the complicated temperature field inside molten pool. The inverse pole figures (IPF) in electron backscatter diffraction (EBSD) analysis suggest that the SLMed 316 L in z-x direction has the preferred orientation in (110) and (111) in x-y direction. Compared with 316 L forgings, SLMed 316 L showed higher average tensile strength, elongation rate and Vickers hardness in 2 directions. The impact toughness was enhanced compared with previous references, which demonstrated that SLM technology is proved an effective way to fabricate FW in the nuclear fusion field.

Conductor design and performance analysis for CFETR magnet

The design work of the China Fusion Engineering Test Reactor (CFETR) was carried out in the past years. The magnet system is composed of 16 toroidal field (TF) coils, 8 central solenoid (CS) coils, and 7 poloidal field (PF) coils. The maximum magnetic field on the TF coils is 14.4 T. Three grades of cable-in-conduit conductors (CICCs), including high-Jc Nb3Sn strand, ITER-like Nb3Sn strand, and NbTi strand, have been chosen for the TF WP winding in high, medium, and low magnetic fields. High-Jc Nb3Sn strand will also be applied in the CS conductor considering the maximum magnetic field, which is higher than the TF coil. For the PF coil, the ITER-like Nb3Sn strand will be applied in Conductor PF1 and Conductor PF7, and NbTi will be applied in others. In this paper, conductor designs for the CFETR magnet including design criteria, choice of conductor configurations, and conductor performance analysis are presented.

Design status of the ECRH system for CFETR

The China Fusion Engineering Test Reactor (CFETR) is the proposed next-generation fusion facility in China. During its two operational phases, a large fraction of externally driven current is foreseen to meet the scientific goals of the two operating scenarios for CFETR. As one of the key design activities of auxiliary heating mix for CFETR, the design of Electron Cyclotron Resonance Heating (ECRH) system is in progress considering the engineering and physical requirements. The conceptual design of the EC system, especially the launcher, will be presented in this paper. Based on the physical requirements and the technical availability, the frequency is chosen to be 170 GHz in order to draw on the technologies development with International Thermonuclear Experimental Reactor (ITER), and the Front Steering Antenna (FSA) is selected as an example to investigate the feasibility of the plant integration design. The beam tracing calculations have been performed to analyze the performance of power deposition and absorption from several launching points. The microwave and optical design of launcher has been carried out to evaluate the maximum power per port and minimize Breeding Blanket (BB) openings.

Evaluation of AC losses in toroidal field magnets in the China Fusion Engineering Test Reactor

Toroidal field (TF) superconducting magnets are the key components of the China Fusion Engineering Test Reactor (CFETR). While the TF magnets operate in a steady-state, they are located in a time-varying poloidal field environment, which may lead to AC losses. In this paper, the AC losses of a CFETR TF magnet during excitation and plasma disruption were analyzed. According to the experimental data, the relationship between the critical current density and the magnetic field (Jc–B) was obtained, allowing the hysteresis loss of a TF magnet to be calculated. The results show that the hysteresis loss during excitation is high, at about 1.26 MJ. The coupling loss of a TF magnet can be evaluated when the time constant of a conductor is obtained theoretically. The AC losses of TF magnet joints were analyzed by the same method, which showed that they are much less than those of a TF magnet.

A new direct external recycling concept for the tritium fuel cycle of CFETR

Tritium self-sustaining operation, which is achieved by a successful closed tritium fuel cycle, is one of the key scientific objectives of China Fusion Engineering Test Reactor (CFETR). The initial tritium start-up inventory and the required TBR are two key parameters which directly affect the possibility of achieving tritium self-sufficiency. In this paper, a new purge gas scheme of He + 0.n% vol. D2 is proposed for tritium extraction of solid breeder blanket, which can avoid the existence of protium in the mixed gas. In this way, the isotope separation system (ISS) in outer fuel cycle (OFC) can be omitted and a new concept of direct external recycling (DER) has been first put forward for OFC, which aims to skip ISS-O and reduce the processing time of OFC; and the feasibility of DER is preliminarily analyzed from four aspects. On this basis, the influences of DER on both the initial start-up inventory and the required TBR have been preliminarily analyzed for different fueling efficiency and tritium burning fraction (TBF) products, and the calculation results are compared with those corresponding to the same direct internal recycling (DIR) fractions. The results indicate that the application of DER can reduce the demand for both the initial start-up inventory and the required TBR, which can improve the possibility of achieving tritium self-sufficiency appropriately. In particular, the amount of “start-up inventory saved by DER” is still attractive under the condition of high fueling efficiency and TBF, and it is in direct proportion to the fusion power. Both of these make DER technology even more attractive for DEMO reactor and future power plant (FPP). These studies can provide some valuable suggestions for the design and optimization of the tritium fuel cycle system of CFETR and FPP.

Numerical Study on Multiple Arcs in a Pyro-Breaker Based on the Black-Box Arc Model

The arc behavior during commutation of a pyro-breaker is the main determinant of performance evaluation. The pyro-breaker discussed in this article is an explosive-driven, extremely fast and non-linear Direct Current circuit breaker. It has been developed for the Quench Protection System (QPS) of superconducting fusion facilities, such as the China Fusion Engineering Test Reactor (CFETR). The feasibility of the Schavemaker differential equation is verified in a simplified 40 kA commutation simulation. The Commutation Section of the pyro-breaker will form multiple gaps after the operation, which causes multiple arc ignitions during the explosion. The influence of the gap quantity on the commutation performance of the pyro-breaker has not previously been studied. A more accurate simulation, which takes the multiple-arc formation into consideration, is proposed and verified under the current of 60 kA. The simulation model, which takes the numerical analysis of the driving mechanism into consideration, will be the designing basis for the pyro-breaker in further development and implementation.

Design and Verification of Calorimeter for CFETR Neutral Beam Injection System Prototype with Negative Ion Source

Currently, a neutral beam injection (NBI) system prototype with a 200-keV negative ion source, called the negative neutral beam injection (NNBI) system, is under construction for the China Fusion Engineering Test Reactor Project. In the NNBI system, the calorimeter is an extremely important high-heat-flux component, and its panels have to withstand a heat flux density of up to dozens of MW/m2 under some extreme conditions; thus, its efficient heat transfer enhancement design and the corresponding supporting structural design are most important, especially under the condition of a 3600-s long pulse. According to the detailed design requirements of the NNBI system, an overall design scheme of the calorimeter based on the heat transfer enhancement structure of the swirl tube (SW for short) is proposed in this paper. By using the gas-liquid two-phase flow boiling model and the supporting structure’s finite element model, the heat transfer performance of the heat exchange module, the mass flow distribution and pressure drop of the entire cooling circuit of the component, and the strength of the support structure are evaluated to verify the feasibility of the design scheme. Finally, based on the proposed design scheme, the detailed design of the temperature monitor system, which has high reliability and economy, is completed. This research provides important theoretical and engineering support for the structural development of the calorimeter for the NNBI verification prototype and will also provide references for the design and development of other internal components of large-scale fusion devices.

Optimization effect of ribs distribution on electromagnetic structural response of CFETR vacuum vessel under MD

In China Fusion Engineering Test Reactor (CFETR), the ribs distribution of vacuum vessel (VV) is a key factor affecting the performance of VV under major disruption (MD) event. In this study, the optimization analysis is carried out to investigate the effect of ribs distribution on CFETR VV under MD event. Besides, the position parameters of the toroidal ribs in equatorial poloidal segment of a VV sector is defined as design variables by degree. The minimizing electromagnetic (EM) force and stress in VV were selected as the objective function. In the optimization process, the EM force and stress in VV were calculated by the EM analysis and structural analysis, respectively. In order to assess the feasibility of optimization analysis, the sensitivity analysis is performed and quantifies the impact of design variables on the objective function. The local sensitivity of the design variables on the EM force and stress in VV ranges from -38.5% to +44.4% and -48.3% to 35.3%, respectively. The optimization results shows that 35.5% reduction of the EM force can be achieved with 18.8% reduction of maximum stress in CFETR VV under MD event. Therefore, the distribution of toroidal ribs has significantly impact on reducing the EM force and stress in VV. The optimization of ribs distribution can improve the performance of the CFETR VV under MD event.