China Fusion Engineering Test Reactor Research Articles

Preliminary design and analysis of the back supporting structure for CFETR WCCB blanket

China Fusion Engineering Test Reactor (CFETR) is a key step to realize the future demonstration fusion power plant and plays an important role in the roadmap of magnetic confinement fusion development in China. As an important in-vessel component, CFETR blanket back supporting structure (BSS) bridges the blanket modules onto the vacuum vessel, withstanding the gravity and electromagnetic loads and distributing the coolant into blanket modules. This paper presents the preliminary design of BSS for the CFETR WCCB blanket and the detailed design of the attachments between BSS and Vacuum Vessel. Based on the proposed design, a global model of BSS and the attachment for one sector has been established with ANSYS code. The thermal-hydraulic analysis has been conducted to assess the heat exchange and temperature distribution. The analyses of individual load conditions and the coupling analysis have been carried out to obtain the stress distribution and deformation. Based on the analyses results, the corresponding design targets and criterion have been met, indicating the feasibility of current BSS and the attachment design. The design and analyses of the BSS together with the attachments can not only provide good reference for the VV design but also provide guidance for design of flexible support used in the blanket pipe system.

Progress of engineering design of CFETR vacuum systems

Vacuum systems for the China Fusion Engineering Test Reactor (CFETR), including pumping system, fueling system, wall conditioning, as well as their control systems, are required to achieve high vacuum, effective plasma fueling, and good wall condition for steady-state plasma operation and to facilitate remote flexible control with high reliability to ensure safe operation. In the last few years, based on the requirements of vacuum systems and subsystems of CFETR, their features and challenges have been analysed, basic parameters have been defined, and preliminary designs have been developed. Herein, nine cryopumps with pumping speed of 68 m³/s for D2 in a molecular flow regime are proposed for use in a torus vacuum chamber to achieve vacuum pressure <1.0 × 10−5 Pa before plasma operation. The throughput of the torus pumping system is designed to be >290 Pa•m³•s−1 to meet the requirement of steady-state operation of CFETR discharge. Pellet injectors and the gas injection system for seeding the fuel and impurity gas and wall conditionings, including baking, GDC, and ICRF-DC, for removal of hydrogen isotopes/impurities of the torus vacuum are also designed. Moreover, integrated design of vacuum systems and space layout of all subsystems are also investigated.

Progress of engineering design of CFETR magnet feeder system

The engineering design project of the China Fusion Engineering Test Reactor (CFETR) was almost done by the end of 2020. The superconducting magnets system of CFETR is significantly different from International Thermonuclear Experimental Reactor (ITER), so as a lifeline of Tokamak system, a new magnet feeder system needs to be designed to meet the CFTER operation requirement. According to the operating requirements of the CFETR magnet system, 23 feeder lines were designed for different coils. In this paper, the progress of the preliminary engineering design of the CFETR magnet feeder system was presented. Combined with the overall design consideration of the CFETR's buildings and other auxiliary systems, the overall layout of feeder system has been completed. The engineering design of main substructures and components were almost completed, including CTB (Coil Terminal Box, 100 kA HTSCL (High Temperature Superconducting Current Lead), CFT (Cryostat Feed-through), ICF (In-Cryostat-Feeder), etc. And some important analysis results were presented to give discussion and conclusion support for the design.

Numerical Analysis of Electron Beam Welding Deformation for the Vacuum Vessel Lower Port Stub of 316L Stainless Steel

The China Fusion Engineering Test Reactor vacuum vessel has strict tolerances requirements and a large number of fully penetrated joints in the manufacturing and assembly processes. As the most complicated component of vacuum vessel, port stubs have stricter tolerance requirements, to guarantee the high assembly accuracy with the vacuum vessel. Due to the small deformation of electron beam welding, this method is adopted for the welding of port stub. However, welding deformation cannot be avoided, thus it is necessary to carry out finite element analysis to analyze and control the welding deformation of the port stub. In this paper, based on thermo-elastic–plastic theory, the double ellipsoid and cone-shaped hybrid heat source can accurately outline the electron beam weld pool contour of the 316L stainless steel with 50 mm thickness, which is further verified by the welding process experiment. The above hybrid heat source is applied to analyze the welding deformation of the lower port stub, and according to the deformation simulation results, the welding fixture is designed to reduce the welding deformation. The welding deformation analysis can effectively provide theoretical data support and practical guidance for the actual electron beam welding process of port stub.

Theoretical study of the Alfven eigenmode stability in CFETR steady state discharges

The aim of this study is to analyze the stability of Alfven eigenmodes (AE) in the China Fusion Engineering Test Reactor (CFETR) plasma for steady state operations. The analysis is done using the gyro-fluid code FAR3d including the effect of the acoustic modes, EP finite Larmor radius damping effects and multiple energetic particle populations. Two high poloidal β scenarios are studied with respect to the location of the internal transport barrier (ITB) at r/a ≈ 0.45 (case A) and r/a ≈ 0.6 (case B). Both operation scenarios show a narrow TAE gap between the inner-middle plasma region and a wide EAE gap all along the plasma radius. The AE stability of CFETR plasmas improves if the ITB is located inwards, case A, showing AEs with lower growth rates with respect to the case B. The AEs growth rate is smaller in the case A because the modes are located in the inner-middle plasma region where the stabilizing effect of the magnetic shear is stronger with respect to the case B. Multiple EP populations effects (NBI driven EP + alpha articles) are negligible for the case A, although the simulations for the case B show a stabilizing effect of the NBI EP on the n = 1 BAE caused by α particles during the thermalization process. If the FLR damping effects are included in the simulations, the growth rate of the EAE/NAE decreases up to 70%, particularly for n > 3 toroidal families. Low n AEs (n < 6) show the largest growth rates. On the other hand, high n modes (n = 6 to 15) are triggered in the frequency range of the NAE, strongly damped by the FLR effects.

Study on the EM loads distribution of the updated WCCB blanket for CFETR during plasma major disruption

As a key component of the China Fusion Engineering Test Reactor (CFETR), the blanket performance directly determines the economic benefit of a fusion device. However, due to the blanket being close to the plasma zone, the electromagnetic loads generated during the plasma disruption play an important role in the safety analysis and structure security. How to accurately and quickly evaluate the safety of the blanket structure is a key issue in the research and design stage. The purpose of this work is to study how to accurately simplify the blanket structure to achieve accurate and efficient calculation of the electromagnetic load distribution in the blanket system. Firstly, the detail structure of the updated CFETR device included the WCCB blanket were introduced. Then, the effect of the inner structure of the blanket and support between the VV and blanket were investigated. The simplified method for the EM analysis model has been verified and an EM result comparison between the detailed model and the simplified model has been performed. It is found that the simplified model can be used in the early design and analysis stages for CFETR blanket design. Through calculation comparison, the supports between the VV and the blanket should be taken into consideration in the EM analysis during plasma disruption. Finally, a fine model including the support structure was considered in the finite element calculation, the results show that the maximum EM force and moment on the blanket modules were generated in blanket #09.

Fatigue life prediction of hypervapotron for CFETR divertor

The engineering design of China Fusion Engineering Testing Reactor (CFETR) will be finished in the end of this year. Compared with Experimental Advanced Superconducting Tokamak (EAST), this is a large development which means closer to commercial fusion reactors. The design objective of CFETR divertor requires a steady-state thermal load of 10–20MW/m2 exhaust. There are several design options for the CFETR divertor. Hypervapotron with its superior performance has been researched for several decades as one important candidate for future fusion device divertor. In this paper, a new design of hypervapotron plasma facing unit (PFU) was developed. Horizontal and longitudinal ribs were adopted to generate more vaporized cores to enhance heat transfer. Some optimizations were performed based on the allowable max heat flux (MAHF) to get the final design. Rensselaer Polytechnic Institute (RPI) boiling model of Eulerian Multiphase fluid and k-epsilon model of Viscous were chosen to deal with sub-cooled boiling. 3Sm rule and Bree-diagram was used to confirm the rationality of the structure after thermal-mechanical analysis. Numerical simulation results suggested that CuCrZr alloy will experience cyclic plastic strain. Based on this, fatigue life prediction of the hypervapotron PFU at a steady-state heat flux of 10MW/m2 were carried out through local stress-strain method, neuber curve and design fatigue life curve. Eventually, a fatigue life of approximately 1.2489 year was verified for CFETR divertor. This paper provides a systematic method for fatigue life prediction and reliability analysis of hypervapotron.

On-line precision control of CFETR multipurpose overload robot using deformation model

The China Fusion Engineering Test Reactor (CFETR) multipurpose overload robot (CMOR) is an important part of the CFETR remote handling (RH) system. It is a type of serial robot with a redundant degree of freedom and heavy loads. Due to its long cantilever and heavy load, the deformation is large during operation. However, the deformation of the CMOR in the RH application cannot be ignored. In this paper, an online precision control algorithm using a deformation model is presented. Using this algorithm, the deformations of the CMOR at a limited number of positions are calculated by finite element analysis and used as training datasets. The joint parameters and loads are used as inputs, and the actual position of the CMOR is the output of the Back Propagation Neural Network (BPNN). The deformation model is trained by the BPNN offline, which can calculate the actual position in the workspace. The deformation of the CMOR can be accurately obtained and rapidly compensated to realize online precision control. During the RH operation in the CFETR, the deformation was updated to 2 Hz and the precision control accuracy was up to 5 mm.

Edge-localized-mode simulation in CFETR steady-state scenario

The EPED1 model and self-consistent core-pedestal coupling in integrated modeling are used to design the pedestal structure of the China Fusion Engineering Testing Reactor (CFETR) steady-state scenario. The key parameters, such as β p and q 95, are based on the grassy edge-localized-mode (ELM) experimental database. In this work, we use the BOUT++ six-field two-fluid code to simulate the onset of the ELM in the CFETR steady-state scenario. The ELM size is around 0.2% in nonlinear simulations, which is in the experimental range of the grassy ELM discharges, 0.1%–1% observed in multiple tokamak devices. Linear and nonlinear simulations show that the dominant high-n ballooning modes peak around n = 40. Compared to type-I ELM crashing dynamics, grassy ELM crashing has a smaller initial crash and is then followed by three phases of turbulence spreading, which are dominated by multi-modes, a high-n mode of n = 45 and low-n mode of n = 5, respectively. In contras to type-I ELM, the perturbation of the high-n mode has a narrow width around ψ = 0.95, and magnetic island formation and reconnection occur only beyond ψ = 0.95, leading to a small initial crash. Mode–mode interaction in the multi-mode coexistence stage stops the growth of individual modes and reduces the transport of particles and heat, and these are the two reasons why the ELM size is small. In–out asymmetry of transient heat flux with a ratio of E out/E in = 3.5 is found during grassy ELM crash. The rise and delay times of the heat flux match the calculation from the free-streaming model. To evaluate the erosion of the divertor target, the energy fluence at the outer divertor target is calculated, which is 0.029 MJ m−2, 5.5 times smaller than the tungsten melting limit 0.16 MJ m−2. The calculated energy fluency still follows the experimental scaling law from type-I ELM experiments. The fluctuation eddies in the toroidal direction show a filament structure at the outer mid-plane. Parallel heat flux patterns with a toroidal mode number n = 10 are found at the outer divertor with an amplitude of 680 MW m−2.

Effect of the Fusion Fuels’ Polarization on Neutron Wall Loading Distribution in CFETR

The China Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, aiming to bridge the gaps between ITER and future fusion power plants. In addition to the temperature dependence, the cross section also depends on the spin states of the reactant nuclei. In this paper, we calculate the neutron source and neutron wall loading (NWL) distributions and investigate the effect of spin polarization on them. For the two unpolarized scenarios at the CFETR, the neutron source distributions have obvious differences, but the poloidal distributions of the NWL have a similar tendency and are just a little different except near the outboard midplane. For the hybrid mode scenario, the maximum of the NWL is near the outboard midplane. However, for the full parallel or antiparallel polarization, the NWL distributions have a big difference in the poloidal direction, and the maximum of the NWL occurs in the upper region of the first wall. The calculation results show that it is possible to optimize blanket design by using polarized fuels at the CFETR, and then increase the working life of the first wall.

Study on electron beam welding process of the first wall of water-cooled ceramic breeder (WCCB) blanket

Water- cooled ceramic breeder (WCCB) blanket is one of the candidate blanket for China Fusion Engineering Test Reactor (CFETR) independently developed by China Institute of plasma physics. The first wall (FW) is a component of WCCB, which directly faces the plasma. FW contains flow channels and has complex shape structure, therefore, it is difficult to complete the full-scale manufacturing by one kind of welding process. Hence, in the development process, the components of FW are first connected by hot isostatic pressing (HIP), and then connected by electron beam welding (EBW), so as to form the full-scaled FW connection. In this paper, the EBW process of FW was studied. The results showed that when the electron beam current was 50 mA and the focus was on the surface, the welded joint with 20 mm penetration and good shape can be formed. After the one time welding and the twice welding (overlapping first welding) of the weld, the tensile test fracture occurred in the base metal. The impact energy of the one time welding was 13 MPa, 18 MPa, and 21 MPa, respectively, and the impact energy of the twice welding was 19 MPa, 20 MPa, and 51 MPa, respectively. The microstructure of the joint was composed of δ-ferrite and lath martensite. When the weld penetration was increased to 50 mm, the lower the focus is, the smaller the porosity at the bottom of the weld is. Similarly, when the lower focus was used, the porosity at the bottom of the weld disappeared basically; The EBW test on the scaled FW with W/Cu sheets showed that when the we took the test pieces with the W-Cu sheet spacing is 4 mm, the welded joint with good surface shape and no internal defects can be formed without scanning deflection electron beam welding, and ultrasonic testing showed that EBW does not affect on the HIP interface. This provides a favorable basis for the future test of equal ratio FW test piece and the welding of actual FW.

First steps of the overall integration design of CFETR

The China Fusion Engineering Test Reactor (CFETR), which is a planned next-step fusion device under engineering design in China now, aims at bridging the gap between ITER and DEMO. As a reactor level device, CFETR will have dozens of complex systems and millions of components and parts. These systems/components shall be consistent with each other in such aspects as function, dimension and interface. The consistency among different systems for CFETR is assured through a dedicated task force of overall integration design, which defines the key engineering parameters of the CFETR tokamak main machine based on the physical design and the CAD rules guiding the mechanical design. The overall integration design also manages the interfaces between the main systems/components in the tokamak machine.

Numerical Analysis of the Convective Heat Transfer Coefficient Enhancement of a Pyro-Breaker Utilized in Superconducting Fusion Facilities

The conductive components of the pyro-breaker in the quench protection system (QPS) have high current density, a large number of electrical contacts and high thermal flux. The water system needs to meet the requirements of cooling and arc extinguishing at the same time. In a previous study, the bottleneck of the steady-state capacity appeared in the barrel conductor of the commutation section, which has a cylindrical cavity. The thermal stability of the commutation section at 100 kA level was simulated in ANSYS/Workbench. The results indicate a certain level of enhancement of the convective heat transfer coefficient of the cavity is required to reach the current capacity. However, the fluid flow inside the cavity is very complex, and the convective heat transfer coefficient is difficult to calculate. In this paper, Computational fluid dynamics (CFD) is applied to the optimization of the cooling water system of the pyro-breaker. By studying the enhancement method of convective heat transfer, optimization of the structure and processing method of the water channel are proposed. The convective heat transfer coefficients of the cylindrical cavity in these optimizations were calculated in CFX. A set of optimizations of the cavity, which can meet the requirements of China Fusion Engineering Test Reactor (CFETR), were obtained and verified by experiments.

Optimization of the cooling design of TF coil case on CFETR

The toroidal field (TF) coil is a crucial system in China Fusion Engineering Test Reactor (CFETR). The operating safety of the magnet will strongly depend on the heat load and the capability of the coolant to remove it. The adequate cooling design of the TF coil case can effectively establish thermal shielding to prevent the heat flux from the TF coil case to the winding. An orthogonal scheme is applied to optimize the cooling design. The effect of the parameters of the cooling channels on the heat transfer from case to conductors is studied, such as the channel numbers N, the distance between the cooling channel and the edge P, the diameter of the pipe D, and the mass flow rate M. Compared with the original design, the optimization design significantly improves the cooling capacity.

Optimization of the gravity support in CFETR vacuum vessel

The China Fusion Engineering Test Reactor (CFETR) is developed as the next tokamak device in China. The gravity support of CFETR VV is designed with a flexible plate structure to allow radial thermal expansion of the vacuum vessel (VV) system and to withstand various design loads. In this paper, the design optimization is developed with Multi-Objective Genetic Algorithm (MOGA) to balance design parameters contributing to stability and strength for the VV gravity support. In this optimization procedure, a parametric FE model is generated to study the structural response of the support corresponding to each design point. Besides, the effects of the design variables on objective function are quantitatively assessed using sensitivity analysis. Through the optimization procedure, optimal design parameters of the VV gravity support are obtained. Compared to the initial design, an improvement of about 50.4 and 401% for the peak stress and buckling load factor of the support plates, respectively.

The Electromagnetic and Structural Analysis of CFETR CSMC Double Current Lead

Before the formal fabrication of central solenoid coil for China Fusion Engineering Test Reactor (CFETR), the R&D of a prototypical coil namely the Central Solenoid Model Coil (CSMC) is performed to accumulate the design and fabrication experiences in the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). As one of the most important auxiliary components, current lead plays the role of transmitting large current from power supply to CSMC winding packs. Of the total 8 sets of current lead, six of them are double current lead. Once CSMC is energized, large electromagnetic force is induced due to the interaction of current flowing in the current lead and back ground magnetic field. And the stress is generated subsequently under the action of electromagnetic force. If the maximum stress is beyond the material strength assessment threshold, the structural design of the double current lead must be redeployed. In order to check the feasibility of the current design of CSMC, the double current lead located in the high field region is selected and the electromagnetic calculation and electromagnetic-structural coupling analysis are carried out subsequently. In the end, the stress in evaluated based on analytical design. The study in the paper will provide reference for the following design optimization of double current lead.

Analysis and Design of a 100 kA HTS Current Lead for CFETR

High temperature superconductors have zero resistance under critical temperature and the HTS tapes have low thermal conductivity. The China Fusion Engineering Test Reactor(CFETR) project requires current leads with current carrying capacity up to 100 kA, making high temperature superconducting current leads the ideal choice. The main work of this paper is the theoretical analysis and optimization design of the 100 kA HTS current lead. Current leads are designed to minimize heat leakage to the cryogenic system and to reduce the mass flow of cooling gas. In this paper, some important design parameters are introduced firstly, and then one dimensional thermal analysis is carried out for the heat exchanger and high temperature superconductor parts. The LOFA time and overheating time of the current lead were calculated through simulation, and it was verified that the design meets the requirements of the project. Finally, the factors that have great influence on the performance of current leads are analyzed. In view of the four influencing factors, which are joint resistance, GHe inlet temperature, hPw and RRR, we respectively produced an optimized design.

Analyses of monoblocks divertor at one-sided high heat flux using RPI wall boiling model

The divertor is one of the core components in the Tokamak magnetic confinement nuclear fusion device, which is responsible for eliminating high-flux energy flow and particle flow. The divertor in China Fusion Engineering Test Reactor (CFETR) is designed to withstand a steady-state thermal load of 10 MW/m2 and a transient thermal load up to 20 MW/m2. The water-cooled scheme is a promising candidate scheme. The model under high heat flux divertor tube system was established by using RPI wall boiling model and the Eulerian multiphase model. On this basis, the fluid-thermal-structure coupling method is used to simulate the complete structure of divertor blocks. In the steady-state condition, three heat transfer patterns are found and compared in the circumferential and axial directions, and methods to enhance heat transfer (including flow velocity and tube structure) are further evaluated. The results reveal the superiority of the screw tube and the swirl tube in the enhancement of heat transfer with high heat flow. For unsteady conditions, the response of temperature and stress for different components are different. The change of flow boundary quickly affects the whole system, while the thermal shock only poses a great threat to external components, and its influence on internal components is lagging and slight.

Safety Analysis of Heat Treatment System and TF Nb3Sn Coil Joint Annealing Experiment

Superconducting magnets for China fusion engineering test reactor (CFETR) are inseparable from Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn coils. The process of “wind-and-react” is used in Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn coil manufacturing. The quality of annealing treatment directly determines the superconductivity of future magnets. In order to complete the heat treatment of TF Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn coil joint, a controlled atmosphere heat treatment system was upgraded by the Institute of Plasma Physics. Heat treatment has the characteristics of complex system structure, long operation cycle, high risk, etc. Therefore, it is necessary to conduct safety analysis and failure prevention of heat treatment. In this paper, the failure mode and effects analysis (FMEA) method will be used for the safety analysis of upgraded system. Through this method, the potential faults of the system can be found, and the faults can be classified into different levels. According to the analysis results, preventive measures for the first-level failure of heat treatment were put forward and the feasibility of preventive measures was verified. The heat treatment of the experimental internal joint of the CFETR TF coil was completed by using the upgraded system, and all the heat treatment technical indexes met the requirements.

A full relativistic model of electron cyclotron current drive efficiency in tokamak plasmas

A fully relativistic model of electron cyclotron current drive (ECCD) efficiency based on the adjoint function techniques is considered. Numerical calculations of the current drive efficiency in a tokamak by using the variational approach are performed. A fully relativistic extension of the variational principle with the modified test function for the Spitzer function with momentum conservation in the electron–electron collision is described in general tokamak geometry. The model developed has generalized that of Marushchenko et al., which is extended for arbitrary temperatures and covers exactly the asymptotic for u≫1 when Zeff≫1, and is suited to ray-tracing and beam-tracing calculations. As a demonstration, the new model is applied in one of widely used ray-tracing codes, TORAY-GA, and yields distinct improvement on the fidelity for the calculation of ECCD in the hybrid operating scenario for the China fusion engineering test reactor.