Engineering Test Reactor Research Articles

Pointing probability Driven semi-analytic Monte Carlo Method (PDMC) – Part I: Global variance reduction for large-scale radiation transport analysis

We propose a Pointing Probability Driven Semi-Analytic Monte Carlo Method (PDMC) to replace the traditional Monte Carlo method for reactor physics analysis and radiation transport analysis. This paper describes the calculation principle and acceleration effect of the PDMC method in radiation transport analysis, highlighting its advantages in dealing with the deep-penetration problem. The PDMC method no longer requires particles to be transported to the detector. During the particle transport process, the global response of each particle state is calculated with a pointing probability to construct global information quickly. The statistical unbiasedness of the global response of a single particle state is demonstrated through rigorous mathematical derivation. The PDMC method does not rely on a deterministic program or iterative calculation, ensuring its universality and efficiency of global variance reduction (GVR). The PDMC method is tested in the China Fusion Engineering Test Reactor (CFETR) and the HBR2 benchmark. The results show that the PDMC method can significantly improve the efficiency of Monte Carlo deep-penetration simulation and is helpful for the GVR for large-scale radiation analysis.

Elevating zero dimensional global scaling predictions to self-consistent theory-based simulations

We have developed an innovative workflow, Stability, Transport, Equilibrium, and Pedestal (STEP)-zero-dimensional (0D), within the OMFIT integrated modeling framework. Through systematic validation against the International Tokamak Physics Activity global H-mode confinement database, we demonstrated that STEP-0D, on average, predicts the energy confinement time with a mean relative error of less than 19%. Moreover, this workflow showed promising potential in predicting plasmas for proposed fusion reactors such as the affordable, robust, compact (ARC) reactor, the European demonstration power plant (EU-DEMO), and the China fusion engineering test reactor (CFETR) indicating moderate H-factors between 0.9 and 1.2. STEP-0D allows theory-based prediction of tokamak scenarios, beginning with 0D quantities. The workflow initiates with the PRO-create module, generating physically consistent plasma profiles and equilibrium using the same 0D quantities as the IPB98(y,2) confinement scaling. This sets the starting point for the STEP module, which further iterates between theory-based physics models of equilibrium, core transport, and pedestal to yield a self-consistent solution. Given these attributes, STEP-0D not only improves the accuracy of predicting plasma performance but also provides a path toward a novel fusion power plant design workflow. When integrated with engineering and costing models within an optimization, this new approach could eliminate the iterative reconciliation between plasma models of varying fidelity. This potential for a more efficient design process underpins STEP-0D's significant contribution to future fusion power plant development.

Study on the effect of neutron shielding design on tritium breeding based on water-cooled ceramic blanket for CFETR

The Chinese Fusion Engineering Test Reactor (CFETR) is a magnetic confinement fusion reactor independently designed and developed by China, which is based on the ITER design experience. As one of the candidates for the CFETR, the water cooled ceramic blanket (WCCB) mainly carries out many important tasks, such as tritium breeding and radiation shielding. The high tritium breeding ability is one of the most significant goals of the blanket. For the design of the CFETR, in addition to meeting the requirements of tritium breeding, the design of the blanket must also consider meeting relevant shielding limits. Due to the limited space of the blanket, the improvement in tritium breeding space will inevitably lead to the reduction in neutron shielding space behind the breeding area, and vice versa. Therefore, under the premise of meeting the requirements of neutron shielding, increasing the tritium breeding space as much as possible is the focus of research. In this work, a three-dimensional neutronics model containing the WCCB blanket is developed, and the neutronics performance is calculated based on 1.5 GW fusion power. A set of nuclear analyses are carried out by the MCNP code, including analysis of the neutron wall load, tritium breeding ratio (TBR), and fast neutron fluence of the TF coil. It is found that the shielding space of certain blanket modules could be optimized. After the shielding optimization, the global TBR increased from 1.168 to 1.186, an increase of 0.018 TBR. The current research has important guiding significance for the future design and optimization of the WCCB for the CFETR.

The effect of triangularity on predicted pedestals for the CFETR

In this work, the dominant unstable magnetohydrodynamic mode and its stable region in the pedestal of the China Fusion Engineering Test Reactor are studied by numerical modeling with the peeling-ballooning theory over a wide range of triangularities (δ) and collisionalities (υ*). A new accessible stable region is found at δ<−0.1 for large βN,ped and υ*. This new stable region expands toward lower βN,ped and υ* with decreasing δ and is totally covered by the peeling unstable region with δ=−0.5 because of the increasing trapped particle fraction (ft,eff). The sensitivity of this new stable region to the kinetic ballooning mode constraint and elongation is studied. For negative and low δ<0.1, the boundary of the first stable region is determined from the ballooning mode. For δ>0.1, the peeling mode becomes dominant as the boundary approaches low s with low υ*, while the ballooning mode is still dominant at the boundary with high υ*. When δ increases beyond 0.46, the first stable region is expanded, and access to the second stable region of the ballooning mode opens up. The entire newly connected region of the first and second stable regions keeps expanding with further increases in δ until δ=0.6. Beyond this value, the ballooning mode becomes more unstable with increasing δ, while the peeling mode is approximately unchanged with increasing δ in this range. The change in the dominant mode and the stable region with increasing δ can be explained by the trade-off between the stabilization effect from the deeper poloidal magnetic well and destabilization due to the enlarged drive term.

Investigation on microstructure and mechanical properties of electron-beam-welded joint of reduced activation ferritic/martensitic steel fabricated by selective laser melting

The microstructure and mechanical properties of the complex low-activation ferritic–martensitic components of the China Fusion Engineering Test Reactor (CFETR) were examined in this study. The quenching–tempering treatment of CLF-1 steel formed by selective laser melting (SLM) (referred to as the QTed material) was carried out to analyze the effects of heat treatment on the microstructural and mechanical properties. Meanwhile, the microstructural and mechanical characteristics of CLF-1 steel electron beam welding (EBW) joints after quenching–tempering (referred to as the EBWed material) were investigated. Compared with the untreated sample (referred to as the SLMed material, with an ultimate tensile strength (UTS) = 979.8 ± 12.7 MPa and elongation (EL) = 3.7% ± 0.4%), the quenching–tempering treatment improved the EL of CLF-1 steel effectively (QTed material, 22.3% ± 0.1%), while it decreased the UTS correspondingly (505.6 ± 11.8 MPa), which mainly was due to the change in the grain size and structure during the heat treatment. The special microstructural distribution characteristics of the CLF-1 fabricated via SLM determined that the fracture mechanism was transgranular ductile/intergranular brittle mixed fracture. After electron beam welding of the QTed material, the tensile properties of the joint (UTS = 710.6 ± 21.4 MPa, EL = 9.7% ± 0.1%) were between those of the SLMed and QTed materials. The fracture appeared in the heat-affected zone (HAZ), proving that the strength of the weld was higher than that of the base metal. The microstructure of the weld was almost completely composed of coarse lath martensite, and the accumulation of dislocation clusters, precipitated M23C6-type carbides, and precipitated MX-type carbides contributed to the improvement of the yield strength (YS). The HAZ could be divided into three regions based on the different heating temperatures, and every region had different grain structures composed of mainly martensite and ferrite. The contributions of four strengthening mechanisms to the YS of CLF-1 steel were analyzed quantitatively, which lays a foundation for further improving the performance of CLF-1 steel and improving its application feasibility on the first wall of a fusion reactor.

Design and thermal-hydraulic analysis for CFETR divertor OVT in consideration of RH compatibility

The divertor is a key in-vessel component of the China Fusion Engineering Test Reactor (CFETR), which is mainly used to remove high heat load and exclude particles. In addition, the CFETR divertor also needs to meet requirements of Remote handling (RH) maintenance due to activation of materials by neutrons. CFETR proposed the concept of hybrid divertor-blanket to meet the requirements mentioned above and improve tritium breeding ratio (TBR). The new concept cancels the cassette and adopts a separate structure. Space below the divertor is occupied by the breeding blanket to improve TBR, and targets are directly fixed on the breeding blanket to improve RH maintenance efficiency. The outer vertical target (OVT) is a high heat load component of the divertor, which needs to remove the heat flux of 20 MW/m2 and neutron radiation heat at 1.5 GW fusion power and meet the requirements of RH maintenance. The structure design, RH compatible structure design, RH process and thermal-hydraulic analysis of OVT are completed in this paper. The relevant design effectively reduces the weight and complexity of RH maintenance and improves RH efficiency of OVT. RH process for OVT proves to have great engineering feasibility. Advanced low activation materials are adopted for OVT. Thermal-hydraulic analysis verifies that OVT can remove heat flux of 20 MW/m2 and neutron radiation heat at 1.5 GW fusion power, which confirms the rationality of the cooling system and effectively prompts the structural analysis of OVT.

Preliminary neutronics/thermal-hydraulics design and analyses of the helium-cooled mixed bed breeder blanket for CFETR under 1.5 GW fusion power

Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device designed with different power levels to validate fusion engineering technologies in China. Helium-cooled ceramic breeder (HCCB) blanket and Water-cooled ceramic breeder (WCCB) blanket are the main candidates. Compared with WCCB, HCCB generally has a higher tritium breeding ratio (TBR) because helium has smaller neutron absorption cross sections than water. However, the TBR of the HCCB decreases with the increase of the fusion power. In this paper, the preliminary neutronics and thermal-hydraulics design of the Helium-Cooled Mixed bed Breeder (HCMB) blanket is proposed. The blanket structure with a fusion power of 1.5 GW is designed based on the current HCCB design of the CFETR. The HCMB adopts helium as the coolant and the mixed Li4SiO4/Be12Ti as the tritium breeder. The neutronics and thermal-hydraulics performances of HCMB blanket have been analyzed in the paper. The results show that the HCMB blanket design has a TBR up to 1.25 with the thermal-hydraulics requirements satisfied.

3D fluid model analysis on the generation of negative hydrogen ions for negative ion source of NBI

A radio-frequency (RF) inductively coupled negative hydrogen ion source (NHIS) has been adopted in the China Fusion Engineering Test Reactor (CFETR) to generate negative hydrogen ions. By incorporating the level-lumping method into a three-dimensional fluid model, the volume production and transportation of H− in the NHIS, which consists of a cylindrical driver region and a rectangular expansion chamber, are investigated self-consistently at a large input power (40 kW) and different pressures (0.3–2.0 Pa). The results indicate that with the increase of pressure, the H− density at the bottom of the expansion region first increases and then decreases. In addition, the effect of the magnetic filter is examined. It is noteworthy that a significant increase in the H− density is observed when the magnetic filter is introduced. As the permanent magnets move towards the driver region, the H− density decreases monotonically and the asymmetry is enhanced. This study contributes to the understanding of H− distribution under various conditions and facilitates the optimization of volume production of negative hydrogen ions in the NHIS.

Study on neutronics modeling with 22.5° model using ANSYS for CFETR

The detailed China Fusion Engineering Test Reactor (CFETR) 22.5° computer aided design (CAD) model is very difficult to convert into Monte Carlo N Particle Transport Code (MCNP). Manually writing MCNP input data is complicated, which is not only time-consuming but also cannot guarantee accuracy. Therefore, in order to improve the efficiency and accuracy of model transformation, modeling with CAD using CATIA is introduced, and MCNP files are converted by ANSYS. This is because ANSYS has a function that converts CAD “stp” format to MCNP input in the geometry section. Meanwhile, ANSYS can also reverse the converted MCNP input file to inspect which module has the problem. Compared with the software platform that can automatically cut, although the CATIA-to-ANSYS method is inferior in terms of automatic operation, it has advantages in accuracy and quickly dealing with error modules. Moreover, it can also perform parametric modeling in CATIA, which facilitates the optimization of the blanket structure. In this paper, the detailed CFETR 22.5° model was developed, and then parametric modeling of the blanket based on CATIA was performed. Finally, a detailed neutronics model is obtained by ANSYS transformation and inspection. Some representative models were initially validated by comparing volume changes before and after conversion. Then, the final neutronics model was used to calculate the nuclear analyses, including the neutron wall loading, fast neutron flux, and nuclear heating on the inboard side. The results show that the volume of the transformed model is basically consistent with the original model, and the error of results is small.

Front-Face RH Compatible Structure Design and Thermal Analysis for CFETR Divertor Dome

The China Fusion Engineering Test Reactor (CFETR) is a device developed to verify the engineering feasibility of a fusion reactor. For CFETR, the divertor is an important plasma-facing component, whose main function is to exclude impurities and remove plasma heat. In addition, the requirement for remote handling (RH) maintenance must be satisfied because of the level of radioactivity in the vacuum vessel after shutdown. The dome is an important component of the divertor, whose main function is to isolate impurity particles as well as to improve the ability of excluding particles. In the optional dome design, a hybrid divertor-blanket concept, a front-face RH compatible structure in plasma-facing units (PFUs), and a RH maintenance scheme for the main bolt are proposed. The vulnerable targets can be replaced directly and thus reduce the RH maintenance time. The dome needs to withstand the heat flux of 10 MW/m2 and nuclear heat in the condition of 1.5 GW of fusion power in the engineering design requirements. Because of the RH compatible structure, higher requirements are demanded for the design of the dome cooling system. In this study, the cooling system and the customized heat transfer structure of dome PFUs are designed to guarantee the maximum heat removal level. The steady-state thermal analysis shows that the cooling system fulfills the design requirements. The concept of the hybrid divertor-blanket and the front-face RH compatible structure for the divertor target have certain reference significance and value for the engineering design and RH maintenance research for the fusion reactor divertor in the future.

Investigation of electron cyclotron wave absorption and current drive in CFETR hybrid scenario plasmas

The investigation of electron cyclotron (EC) wave absorption and current drive has been performed for the China Fusion Engineering Test Reactor (CFETR) hybrid scenarios using the TORAY code. To achieve the physics goal of the EC system in CFETR, a total of four wave frequency values and nine locations of launching antennas have been considered, and the injection poloidal and toroidal angles have been scanned systematically. The electron cyclotron current drive (ECCD) efficiency of the 170 GHz EC system is quite low due to the wave-particle interactions being located at the low-field side. To optimize the ECCD efficiency, the wave frequency is increased up to 221–250 GHz, which leads to the power being deposited at the high-field side. The off-axis ECCD efficiency can be significantly enhanced by launching EC waves from the top window and injecting them towards the high-field side. The optimized ECCD efficiency at ρ = 0.32 and at ρ = 0.4 is 2.9 and 2.2 times that of 170 GHz, respectively.

Structural design and strength analysis of the calorimeter pipeline joint for the negative neutral beam injection system of China fusion engineering test reactor

The negative ion-based neutral beam injection (NNBI) verification prototype system is one of the core components of China Fusion Engineering Test Reactor (CFETR). To effectively cut off the beam and debug the neutral beam device, the power measurement system plays an indispensable role. In this study, we examine the detailed design of the pipeline joint structure of the calorimeter verification prototype, focusing on the impact of the structural design of the target panel pipeline joint on the structural strength and manufacturing process required by the cooling circular pipeline joint of the calorimeter. Based on this, the manufacturing process and strength of the integrated pipeline structure of the target panel are inspected under a design pressure of 4 MPa, including theoretical analysis of the integrated pipeline pressure-bearing structural design, welding technology design, and finite element simulation under high pressure. Accordingly, the feasibility of the final manufacturing scheme and key technologies are analyzed. The pressure test of the sample under an internal N2 pressure of 4 MPa verifies that its structure can meet the design requirements, thereby providing reference for the effective structural design and manufacturing of the NNBI calorimeter prototype.

Improved multiphysics model of the High Temperature Engineering Test Reactor for the simulation of loss-of-forced-cooling experiments

We present a multiphysics model of the High Temperature Engineering Test Reactor for comparison with past and predict future loss-of-forced-cooling (LOFC) experiments. The approach selected combines (1) 3-D full-core superhomogenization-corrected neutronics, (2) 3-D full-core homogenized or semi-heterogeneous heat transfer (macroscale), (3) 2-D axisymmetric fuel rod heat transfer (pin-scale), and (4) 1-D thermal-hydraulics channels. Although large uncertainties remain, the time and magnitude of the first power peak after re-criticality is predicted within 1.5 h and 175 kW, respectively. The novelty of our work includes (1) a new macroscale/pin-scale heat transfer coupling approach relying on gap conductance to drastically speed up numerical convergence by two orders of magnitude, (2) determination of a radial effective thermal conductivity, reproducing the semi-heterogeneous re-criticality time within one hour using a homogenized macroscale model, and (3) a preliminary study of the reactor’s early behavior following a LOFC event, enabling further assessment of numerical models against fission power measurements.

Solving redundant inverse kinematics of CMOR based on chaos-driven particle swarm optimization algorithm

The China Fusion Engineering Test Reactor (CFETR) multipurpose overload robot (CMOR) is used to maintain the vacuum chamber inner components damaged by heat loads, electromagnetic fields, and nuclear radiation. The CMOR is a redundant robot whose configuration does not meet the condition of the Pieper criterion. Redundant kinematic problems can only be solved by numerical methods, not analytical methods. Conventional numerical iterative methods include considerable computational load, accumulated errors, and the Jacobian matrix's singularity. To solve redundant inverse kinematics efficiently, we compared the behaviors of 25 versions of particle swarm optimization (PSO) algorithms with 12 one-dimensional chaotic maps under unimodal and multimodal test functions. Moreover, we selected three chaos-driven PSO algorithms with optimal convergence performance to address CMOR inverse kinematics. The experimental results indicated that chaos-driven PSO algorithms have higher computational efficiency and can effectively improve the speed and accuracy of algorithm convergence. This proposed algorithm delivers a novel and efficient method for inverse kinematics of redundant robots based on chaotic maps.

Investigating the Manufacturing Process of Heat-Treated Nb3Sn-CICC Conductor for CFETR CSMC Modules

The center solenoid model coil (CSMC) of the China fusion engineering test reactor (CFETR) requires a high heat treatment temperature of 650 °C, which would destroy the insulation layer by carbonization and cause coil failure. To avoid this, using the manufacturing technology and routine process of the ITER center solenoid coil as a reference, in this article, a solution is introduced that is applied for manufacturing the CFETR CSMC module. The coil is heat treated before insulation and after winding. A single turn of the heat-treated coil is separated and the wrapping turn insulation is applied, and then the separated layers are reversed back into the integrated coil. To manufacture the integrated coils, the heat-treated coils are molded. After heat treatment, the interlayer separation distance of the Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn cable-in-conduit conductor in the coil should satisfy the allowable strain range of the superconductor, which is crucial for designing the rotational support structure of the interlayer separation. In addition, to achieve the accurate final shape of the coil and meet the design requirements, as well as to ensure coil contour accuracy after layer by layer separation, the antideformation ability of the coil mold is a key research topic. Herein, mechanical properties of the key components were optimized and numerical calculations and ANSYS simulations were used for statistical analyses. Results from the analyses indicated that the equivalent stress on the device is less than the maximum permissible stress of the raw material. The device was manufactured and compared with a design mold. The overall deformation error was <5%, movement error of the electric frame was <2%, rotation error of rotating was less than ±1° per 360°, levelness between the upper and lower platform was <1 mm, and motion error was less than ±1 mm per 100 mm. These results demonstrated the optimized structure design of the device. In the as-obtained device, the coil could separate a single turn and wrapping turn insulation could be applied. Parameters of the manufactured coil satisfied all the CSMC design requirements; these results indicate that the proposed device is suitable for manufacturing the CFETR CSMC module coil.

Design and Analysis of the Inlet Valve for the CFETR Torus Cryopump

The China Fusion Engineering Test Reactor (CFETR), a superconducting magnetic confinement tokamak fusion reactor, will develop a high-performance torus cryopump to pump torus plasma exhaust gas. The inlet valve is one of the key components of the cryopump, and it is used to isolate the cryopump from the plasma for regeneration, to control the pumping speed of the cryopump, and to operate as a pressure relief valve in case of a failure, such as the cryopipe breaking inside the cryopump chamber. This paper presents a novel inlet valve. Ensuring that the design of the inlet valve meets the above requirements will be a challenge. In order to verify the reliability of the inlet valve, its critical components are analyzed and optimized by the Finite Element Method. The effect of the stroke of the inlet valve on pumping performance is then estimated by the Monte Carlo Method, and the pressure profile in the whole flow field is studied to predict the cryopump’s behavior. Finally, the seismic capacity of the optimized inlet valve is analyzed, and the mechanical performance of the inlet valve is shown to meet CFETR design criteria. These design and analysis results will provide technical support and references for the development of the CFETR torus cryopump.

Variable-curvature elephant trunk robot in nuclear industry

In this paper, a variable curvature Elephant trunk robot (ETR) is designed based on the bionic principle to complete the visual inspection and assisted flaw detection tasks in the bottom of the divertor and the vertical port complex pipe area of the China Fusion Engineering Test Reactor (CFETR). It offers a miniaturized and lightweight structure with similar load capacity and position accuracy to a rigid robot. It can be installed as a sub-system of the CFETR remote handling maintenance system on various mobile maintenance platforms such as the CFETR multipurpose overload robot and can penetrate various confined spaces and multi-obstacle environments to complete operational tasks. The kinematic model of the ETR is established based on the Denavit Hartenberg (DH) method. A prototype is built according to the design method and trajectory tests are carried out to verify the rationality of the structural design. For ETR trajectory control requirements, three different trajectory tracking control methods are proposed in this paper. The effectiveness of the trajectory tracking methods is verified by simulation. The Elephant Trunk Robot design and control strategy provides a reference for narrow space maintenance task in the CFETR.

Evaluation of the mechanical properties of CICC jacket for CFETR superconducting magnet

Modified 316LN and JK2LB stainless steel were selected as candidate jacket materials for superconducting cable in conduit conductors (CICC) of the China Fusion Engineering Test Reactor (CFETR). Different superconducting conductors need to undergo different cold working deformations and different heat treatment processes during the manufacturing process. Therefore, a systematic investigation was done to examine the influence of different cold working deformations and different heat treatment processes on the mechanical properties of the two materials. The mechanical properties of materials were measured at 4.2 K, 77 K and 300 K, and Young's modulus, yield strength (0.2% offset), ultimate tensile strength and elongation at failure were reported. In addition, the fracture morphology of all tensile samples was observed by backscattered-electron imaging (BSEI). The results show that the effect of increasing cold deformation of extrusion to improve the yield strength of the jacket is not obvious. In addition, the heat treatment schemes of different superconducting materials have limited influence on the performance of the jacket. These two materials have the potential to be used in CICC conductors prepared from different superconducting materials.

Analyses and optimization of the CFETR power conversion system with a new supercritical CO2 Brayton cycle

In this paper, the Chinese Fusion Engineering Testing Reactor (CFETR) power conversion system, with a supercritical CO2 (SCO2) Brayton cycle, is designed, analyzed and optimized. Considering the pulse operation of the reactor, a heat storage loop with high temperature molten salt and low temperature concrete is introduced. Based on the parameters of the first cooling loop, the CFETR power conversion loop is designed and studied. A new SCO2 Brayton cycle for the CFETR dual heat sources, blanket and divertor, is developed and optimized using a genetic algorithm. Compared to other simple and recompression cycles, it is shown that the new SCO2 Brayton cycle combines maximum thermal efficiency with simplicity. Exergy analyses are carried out and show that the exergy destruction rates of turbine and heat exchangers between different loops are the largest due to the large turbine power and the large temperature difference. The exergoeconomic analyses show that the fusion reactor accounts for the main cost, which is the key to the economy of fusion power generation. The following sensitivity analyses show that the hot molten salt temperature has a major influence on the system performance. Finally, several multi-criteria optimization algorithms are introduced to simultaneously optimize the three fitness functions, the cycle thermal efficiency, the system exergy efficiency and the total system product unit cost. It is found that the maximum thermal efficiency, the maximum exergy efficiency and the lowest total system product unit cost can be obtained almost simultaneously for the new CFETR power conversion system, and this optimal operation scheme is presented.

DC active power filter based on model predictive control for DC bus overvoltage suppression of accelerator grid power supply

The China Fusion Engineering Test Reactor plans to build a 200 kV/25 A acceleration grid power supply (AGPS) for the negative-ion-based neutral beam injector prototype system. The AGPS uses a rectifier-inverter-isolated step-up structure. There is a DC bus between the rectifier and the inverter. In order to limit DC bus voltage ripple and transient fluctuations, a large number of capacitors are used, which degrades the reliability of the power supply and occupies a large amount of space. This work finds that due to the difference in the turn-off time of the rectifier and the inverter, the capacitance mainly depends on the rectifier current when the inverter is turned off. On this basis, an active power filter (APF) scheme is proposed to absorb the current. To enhance the dynamic response ability of the APF, model predictive control is adopted. In this paper, the circuit structure of the APF is introduced, the prediction model is deduced, the corresponding control strategy and signal detection method are proposed, and the simulation and experimental results show that APF can track the transient current of the DC bus and reduce the voltage fluctuation significantly.