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The rotor/stator interaction noise is the primary noise source of modern turbofan engines. Vane unsteady pressure test is essential to understand the generation mechanism and propagation characteristics of the noise, as well as examine the passive/active noise control technologies. However, limited by the Vane’s thickness as well as the requirement for high spatial and temporal resolutions of the sensor, traditional transducers still have considerable limitations and difficulties carrying out the measurement. In this paper, a polyvinylidene fluoride (PVDF) piezoelectric-film sensor array was developed for the vane unsteady pressure test. The dynamic characteristics of the sensor were obtained by the shock tube test. The calibration results further confirm the effectiveness of the self-developed PVDF sensor and also show the advantage in frequency response over traditional pressure transducer. Time histories of the vane fluctuating pressures were obtained by the PVDF sensor. The blade passing frequency (BPF) harmonic component and the broadband component of the unsteady pressure were analyzed individually by time-domain averaging and spectral analysis. The distributions of BPF content and broadband content on both sides of the vane at different fan tip speeds and operating conditions were investigated in detail. The results indicate that the operating conditions have little effect on both of them. The broadband component yielded by the rotor-wake turbulence seems to be quasi-isotropic. The average broadband level increases linearly with the fan tip speed. The developed PVDF sensor shows potential in noise evaluation and noise source diagnosis, which can properly guide further noise reduction technology development.

An accurate life prediction is important to the design of a high-speed rotary blade subjected to multilevel cyclic loads. The widely used Miner’s rule and uniaxial stress prediction method always deviate from actual life of the blade. A prediction method based on Chaboche’s nonlinear damage evolution model is utilized to predict the multilevel cyclic fatigue life of a compressor blade subjected to start-up centrifugal force and working aerodynamic force. Chaboche’s model is verified by comparing with experimental data of different materials. The blade life predicted by Chaboche’s rule and Miner’s rule are compared, and it is found that Miner’s rule might overestimate the blade life under the typical loading spectrum of start-up centrifugal force and working aerodynamic force. To study the impact of multiaxial stress state on the blade life, the life predicted by uniaxial stress method is compared to that predicted by multiaxial stress method, and it demonstrates that the multiaxial stress state of the blade should not be neglected. Finally, the crack propagation of the blade under multiaxial fatigue loads is simulated successfully by element deletion technique, which is conducted by translating Chaboche’s multiaxial model into a user defined UMAT program in ABAQUS. The predicted crack propagation life is compared with that predicted by an approximate Paris law method plate model. This research proves that the method to predict the blade life subjected to multilevel cyclic loads based on multiaxial Chaboche’s model could provide a valuable reference for engineering blade fatigue design.

The calculation of the real contact area between rough surfaces is an important prerequisite for studying the contact characteristics of mechanical joints. However, it is challenging to accurately calculate the real contact area between rough surfaces by existing theoretical and experimental methods. Therefore, an efficient and practical equivalent calculation method for the real contact area between three-dimensional(3D) rough surfaces based on two-dimensional(2D) profiles is presented in this paper. The main advantage of the presented method as compared with those available in the literature is that the real contact area can be predicted only by the section of the 3D rough surfaces in contact, which greatly reduces the computational cost. The sampling strategy of sections is studied to improve the accuracy of the proposed method. The accuracy of the proposed equivalent calculation method is verified by numerical simulation. The results show that the proposed equivalent calculation method can efficiently predict the real contact area between rough surfaces within the scope of engineering application.

Most of the computer codes for nuclear power plants safety analysis were developed based on conservative methods, which would lead to unnecessary excessive margins and more difficulty for nuclear power plants design and improvement. Using best-estimate plus uncertainty analysis method may reduce or eliminate these unnecessary limits. The NRC had used the code as a basis for safety reviews of the AP series nuclear power plants, driving the development of the best-estimate plus uncertainty analysis methods (BEPU) in the independent audit calculation. It had become an international trend to use BEPU in accident analysis and safety review of nuclear power plants. At present, the calculation method of the best-estimate in the world is used to better understand the accident sequence in the water-cooled reactor, and uncertainty assessment is the necessary supplement to the best-estimate method. The research and application of BEPU method are becoming more and more extensive. Best-estimate method was combined with best-estimation program and the initial and boundary conditions. It accurately described the current safety margin. It was required to consider each impact analysis of the available information and data, including the best-estimate model, the uncertainty calculated in the transient state, the input and the uncertainty of input parameters. Compared with the conservative evaluation method, the BEPU analysis method defined the gap between the calculated result and the real value through the uncertainty analysis, and made a more reasonable evaluation of the safety margin. In the process of nuclear power plant safety review, National Nuclear Safety Administration (NNSA) had adopted the best-estimate plus uncertainty analysis method to review the LOCA accident. In this paper, the four best-estimate plus uncertainty analysis methods are introduced, the source of uncertainty and uncertainty statistic methods are described. Then, the application of best-estimate plus uncertainty analysis method in nuclear safety evaluation are discussed.

The reliability and safety of aero-engine are often the decisive factors for the safe and reliable flight of commercial aircraft. Hence, the vibration source location and fault diagnosis of aero-engine are of prime importance to detect faults and carry out fast and effective maintenance in time. However, the vibration signals collected by the sensors arranged on the casing of the aero-engine are generally the mixed signals of the main vibration sources inside the engine, and the components are extremely complicated. Therefore, the vibration source identification is a big challenge for a fault diagnosis and health management of the engine. In order to separate the key vibration sources of rotating machinery such as aero-engine, a Joint Wavelet Transform and Time Synchronous Averaging based algorithm (JWTS) is proposed in this paper. Based on the fact that the fundamental frequency and its harmonic and sub-harmonic components are generally included in the vibration spectrum of shaft fault signal of rotating machinery, wavelet transform and time synchronous averaging algorithm are combined to extract them. The algorithm completes separating the main vibration sources with three steps. First, the source number and fundamental frequency of each source are estimated using the wavelet transform. Second, every source is extracted from each observed signal by the time synchronous averaging method. Time synchronous averaging method can effectively extract a signal of cycle and harmonic rotor components and can suppress noise. Third, the optimal estimation of each source is determined according to signal’s 2-norm. Since the extracted source with a larger energy is closer to the real source, and signal’s 2-norm is a good indicator of the signal energy. Hence, the key vibration sources related to rotary speeds of the engine are obtained separately. The method is verified by synthetic mixed signals first. Three periodic signals of different frequencies are used to simulate the vibration sources of the aeroengine. The fundamental, harmonic and sub-harmonic components of them, as well as Gaussian white noise, are randomly mixed. The results show that the JWTS algorithm can estimate the number of the main sources and can extract each source effectively. Then the method is demonstrated using vibration signals of a real aero-engine. The results indicate that the proposed JWTS method has extracted and located the main sources within the aero-engine, including sources from the low-pressure rotor, high-pressure rotor, combustion chamber and accessory. Therefore, the proposed method provides a new fault diagnosis technology for rotating machinery, especially for a real aero-engine.

The acquisition of the oil film pressure and forces on the bearing pads through experimentation is crucial to understanding the characteristics of journal bearing. Lots of efforts had been taken in film pressure measurement, and the pressure was obtained at specified position on the bearing pads. However, due to the space and structure constraint, merely very limited number of the point pressure can be obtained with traditional sensors and acquiring the detail pressure field on whole bearing pad surface is still an open challenge. In this paper, a method based on thin-film sensors technique is proposed and employed to measure the pressure distribution and oil film forces of journal bearings. The measurement is conducted on a cylindrical journal bearing with two axial grooves, and ElectroMechanical Film sensor arrays are designed and laid on the surface of the bearing pads. The oil film pressure is acquired at up to 32 measurement points in total along the bearing pads in both circumference and axial directions. The pressure distribution in a wide rotation speed range is obtained successfully by using fitting algorithm. Furthermore, the oil film forces on horizon and vertical direction are obtained through the integration of the measured pressure filed. The test results prove that it is feasible to measure the oil pressure filed of journal bearings using ElectroMechanical Film piezo-film sensor array.

The position of fluid film in journal bearing will change while the journal moving in bearing, which can be named fluid film boundary migration (FFBM). It is usually ignored in the calculation of linear dynamic coefficients. While, the errors brought by this neglection was not ever investigated in detail. In this paper, the influence of FFBM on bearing dynamic coefficients and rotor system dynamic behaviors are investigated. A new perturbation-based model is proposed to take the FFBM into account by modifying the boundary conditions of governing equations. It is then verified by the experimental results and analytical results from previous research. Furthermore, the effects of FFBM on stiffness and damping in two typical journal bearings are investigated. The result indicates that the FFBM has a significant influence on dynamic coefficients of full circular journal bearing but little impact on journal bearing with axial grooves. Moreover, it affects the stiffness and damping more significantly in the cases of large length-to-diameter ratios or small eccentricity ratios in full circle bearing. Finally, the dynamical behavior of a rotor-bearing system with considering the FFBM is also investigated. The result shows that the FFBM of oil film has remarkable influences on the instability threshold and imbalance responses of the rotor system, which should not be ignored. The conclusions obtained in this research are expected to be helpful for the design of full circular journal bearings or rotor-bearing systems.

Experimentation is an effective approach to study the dynamic coefficients of journal bearings because experimental results are more reliable than theoretical calculations. Dynamic coefficients ide...

This paper presents a contact model to describe the normal and tangential contact behaviors of the rough interface with multiscale topographies. The coupling effect of macroscopic profile deviation and mesoscopic roughness on the contact behaviors is taken into account. First, an improved force expression of single contacting asperity is proposed to characterize the continuous and monotonic feature in the whole deformation process. By adopting the statistical summation, the contact model for the surface with microscopic topography is established. Second, the mesoscopic waviness deformation is characterized with microscopic asperities through multiscale modeling. The macroscopic geometric deviation is defined as a linear shape function. By combining the macro-, meso-- and microscopic topographies, the multiscale normal contact model is established. Then, in terms of tangential model, the Jenkins element is applied to express the slip-stick behavior of asperity. Based on the normal contact model and Coulomb law, the distribution function of the yield slip force is obtained to establish the tangential contact model. Furthermore, the predicted contact responses are compared with experiments and finite element model to validate the effectiveness of the proposed model. Finally, the influences of the multiscale topographies are analyzed, presenting the variation tendencies of contact responses. This work provides a feasible and effective way for studying the contact response of rough interface with multiscale surface topographies.

The semi-analytical model (a.k.a. the tube-in-channel model) needs less empirical inputs and considers interactions between tube movements and cross-flow variables, and is well used in fluidelastic instability (FEI) threshold prediction of tube bundles. However, despite numerous flexible tubes in engineering practice, tube-in-channel models with more than 7 tubes are not yet studied. In this paper, a new tube-in-channel model with a layout of 14 flexible tubes, rather than the reported 1, 5, or 7 tube(s), is established and used for FEI threshold prediction of rotated triangular tube bundles in two-phase cross-flow. Five sets of experimental data are used for verifying this model, of which the accuracy and efficiency are compared with those of tube-in-channel models with 1 and 7 tube(s). Influences of the attachment angle, the separation angle, and the time-delay parameter on performances of these three models are also investigated. Since its numerical predictions agree well with the experiments, the present 14-tubes model shows potential applications in steam generator design.

The asymmetric pressure wave caused by Loss of Coolant Accident (LOCA) transients in multi-loop type Pressurized Water Reactor (PWR) leads to great shock to the reactor core barrel. The instantaneous shock causes severe damage to the reactor vessel structure integrity. In this paper, a new method for reactor core barrel instantaneous force calculation was proposed based on the widely accepted reactor system analysis code RELAP5. The reliability of RELAP5 code for pipe blow down and pressure wave propagation simulation was assessed against Edward pipe problem firstly. Then the typical three-loop PWR system model was established utilizing Symbolic Nuclear Analysis Package (SNAP) software and the small break LOCA scenarios with different break diameters and break locations were analyzed. The maximum instantaneous asymmetric force applied on the reactor core barrel was achieved under various break conditions, providing the boundary conditions for the core barrel stress distribution and deformation calculation. Results show that the cold leg break LOCA transient leads to great instantaneous asymmetric force, while the hot leg break LOCA transient has no obvious asymmetric force generated. The core barrel instantaneous stress concentration and deformation were discovered. This research could contribute to the reliability assessment of PWR core barrel under LOCA scenarios.

Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is coating a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Through considering the loading patterns, loading amount and self-shielding effect and so on, when the outer diameter of MA is 0.520 cm, the transmutation rate of MA is 38.5%, it is the optimal configuration, firstly transmutation rate and loading amount of MA with the pattern of MA in the water gap of burnable poison rods are larger than ones with the pattern of MA substitute for part burnable poison. Secondly the highest transmutation rate means the least loading amount of MA and the lowest transmutation rate means the most loading amount of MA. In order to keep reasonable transmutation rate and loading amount, when the outer diameter of MA is 0.520 cm, it would not reduce dramatically the amount of MA and the MA loading of this loading pattern is equivalent to the discharge amount of five pressurized water reactor, at the same time, the transmutation rate can keep reasonable level.

A systematic uncertainty analysis strategy based on stochastic sampling method is developed for best estimate analysis of the rod ejection accident in pressurized water reactor (PWR). Self-developed SAMP module and DAKOTA code are used to randomly sample the multigroup nuclear data, important neutronics and thermal-hydraulic (T-H) input parameters respectively. The 3D coupled reactor core model for Almaraz PWR plant is established using RELAP5/PARCS code system and the uncertainties of the core power, fuel enthalpy and local power distribution are obtained. The results indicate that the uncertainties of nuclear data and input parameters can result in large uncertainty on the results which should be carefully considered.

Aqueous Homogeneous Solution Reactor concept has been proposed for producing medical isotopes(Medical Isotope Production Reactor-MIPR). However, there are several difficulties in transient calculation of aqueous homogeneous solution reactors. First, there are no assemblies in the core which is different from the traditional reactor core. Second, the operation of aqueous solution reactor at a power of 200kW will generate radiolytic-gas bubbles. The void volume created by these bubbles in the solution core will introduce a strong negative reactivity feedback. Third, the complex structure of the coolant pipes immersed in fuel solution requires unstructured neutron diffusion calculation methods. Therefore, analytic basis functions expansion nodal method for arbitrary triangular-z node is established to solve the complex structure geometry neutron diffusion equation. Based on this, a software named TABFEN-K has been developed to solve the three-dimensional space-time neutron kinetic equations. Then，TABFEN-K code is used for typical accident analysis of a solution reactor. A simplified geometry model, bubbles generation model , thermal conduction model and and cross section feedback model are given in this paper. A software called TABFEN-MIPR is developed and used for the simulations of the control rod ejection and drop. The same characteristics in the transient process with the results from literatures are obtained.

Spline couplings which have simple structure, high reliability and can compensate torque transmission error are widely used in rotating machineries, such as aeroengine and gasturbine, etc. Recent efforts show that it is potential to make the rotor system losing its stability. Nevertheless, the experimental study of rotor system with spline coupling is rare and inadequate. This indicates a need to study the factors that affect the stability of rotor system with spline coupling experimentally. In this paper, a specially designed spline connection rotor test rig has been built and used to simulate a multi rotor system of turboshaft engine. The experimental instability characteristics of spline connected rotor system are presented. The instability speed and critical speed under different conditions such as lubrication conditions, external damping, load torque, spline tooth error and fit type of internal and external spline are measured. Based on the above-mentioned results, the effect rules of the influence factors on spline connected rotor system stability are studied. Results show that lubrication can effectively weaken the vibration of the system. The increased external damping makes the stability better when the spline coupling is unlubricated. With the increasing of load, the subharmonic vibration decreases after the system loses its stability, the system stability becomes better. The stability of spline coupled system with larger tooth error is better than that with normal one. Normal fit-up spline coupling improves the system stability under the conditions of lubrication and small external damping. This study may be helpful to get the favorable parameter setting of spline connected rotor system for avoiding instability and reducing vibration.

AP1000 reactor is the typical “third-generation” nuclear power plant in the world at present. The primary system of AP1000 nuclear power plant was modeled using RELAP5/MOD 3.3 code, and the transient thermal hydraulic characteristics were analyzed under complete loss of flow accident (CLOFA). The calculation results by RELAP5 code were compared with those of the LOFTRAN codes. The research results show that the RELAP5 model can accurately simulate the transient thermal hydraulic characteristics of AP1000 under the accident of complete loss of forced reactor coolant flow. Meanwhile, comparison was made between the results of complete loss of reactor coolant flow initiated by loss of voltage and frequency reduction of offsite power, and the analysis indicates that the results of complete loss of flow initiated by complete loss of voltage bound the complete loss of flow initiated by reduction in the reactor coolant pump motor supply frequency with a frequency decay of up to 5 hertz per second.

Based on the PWR best estimate program RELAP5/MOD3.3, the reactor coolant system and the passive core cooling system of AP1000 Nuclear Power Plant were modeled and analyzed. Some key transient parameters were obtained, including RCS pressure, break discharge, etc. The result is in close agreement with the result generated by NOTRUMP code from Westinghouse USA. The calculation results show that safeguard system of APl000 can effectively reduce the pressure of the first loop, also can prevent core uncovered. The safety of AP1000 during a small break Loss of Coolant Accident caused by DVI line breaking is verified.

AP1000 nuclear power plant model is founded via RELAP5/mod 3.3 program in detail. Refer to SGTR accident basic assumptions in the safety analysis report, calculations for AP1000 nuclear power plant SGTR accident are carried out. And through comparison between the result got and the result from safety analysis report calculated by LOFTTR2 program, the two results are very consistent. It's found that even if the operators do not interfere, AP1000 nuclear power plant protection systems and passive design measures will trigger an automatic response measures, it can terminate the SG heat transfer tube leaking and make RCS be stable in a safe condition and to prevent SG overflowing and ADS actions occurring.

The unsteady flow in an axial flow compressor rotor is responsible for noise generation and unexpected excitation of blade vibration; therefore, compressors may suffer from fatigue failure of blades and eventually become unable to operate. At present, a set of Kulite pressure transducers mounted on the casing wall is widely used to acquire the pressure field of unsteady flow. However, it will affect the compressor structure and it is physically impossible to mount at a small spacing to capture the variation of casing wall pressure precisely because of the size of transducers. For the requirement of measurement, a self-developed PVDF (polyvinylidene fluoride) piezoelectric-film sensor array measurement system was applied to measure and capture the pressure field distribution over the compressor rotor blade tip clearance. In this system, a multi-measuring points PVDF array, in which 40 measuring points was contained in a 70mm by 40mm rectangle area, was attached on the casing wall of the rotor blade tip. This new technique has great advantages over traditional measurement technique, such as wide frequency response, high space resolution, less impact on the compressor, easy installation and so on. Based on a multistage axial compressor rig test, the unsteady flow pattern on the casing wall was well captured by the PVDF array. And it is revealed in all measuring points that the resonance occurs all over the compressor is under one single frequency, and the pressure level of the acoustic resonance exceeds the level of the usually dominated blade passing frequency when close to the stability limit. At the same time, the acoustic resonance will be modulated with the blade passing frequency.

Best-estimate codes have become more attractive to people who are dedicated to reactor safety analysis. With the purpose of getting more realistic results and larger safety margins, best-estimate plus uncertainty (BEPU) analysis method was developed to replace conservative safety analysis method. For example, ASTRUM (Automated Statistical Treatment of Uncertainty Method) was developed to analyze AP1000 Large Break Loss of Coolant Accident (LB-LOCA). This paper presents a LB-LOCA BEPU analysis by means of thermal hydraulic code RELAP5 coupled with DAKOTA. Both these two codes are coupled and invoked by Symbolic Nuclear Analysis Package (SNAP). The uncertainty parameters are selected according to LB-LOCA phenomena identification and ranking table. The corresponding distributions assigned to each parameter have been extracted from different works in open literature. With Wilks’ formula, 59 cases were calculated to ensure 95 % probability coverage at 95 % confidence level. The result shows that the Advanced Large PWR can meet the acceptance criteria of LB-LOCA analysis with enough margins.

In this paper Hualong One nuclear power plant model is established using TRACE code in detail. First according to analysis assumptions in PSAR, research of complete loss of flow accident (CLOFA) for Hualong One is carried out, and then research of core DNBR through the accident is carried out using FLICA III-F code. By comparison the research results and PSAR results are very consistent, it means the research method is feasible and the results are credible. On this basis, the more conservative analysis assumptions are chosen, further accident analysis is carried out. It found that all the research results can meet the acceptance criteria of CLOFA. It indicates that Hualong One is very safe and reliable at CLOFA.

The influence of mass-flow-rate ratio of inner to outer secondary air on gas–particle flow characteristics was determined in the near-burner region of a centrally fuel-rich swirl coal combustion burner. Velocity and particle volume flux profiles and normalized particle number concentrations were obtained. Peaks in tangential mean velocity and three-dimensional root-mean-square fluctuation velocities were found to decrease as the mass-flow-rate ratio increased. Moreover, the peaks in the mean axial velocities and particle volume flux near the wall increased, whereas those near the chamber axis decreased. Simultaneously, both recirculation zone and swirl number decreased as the mass-flow-rate ratio increased.

In this paper numerical simulation of PRHR HX and IRWST is demonstrated using FLUENT, and different numbers of C-type heat transfer tubes and coolant inlet temperature’s effects for the residual heat removal capacity of PRHR HX, IRWST thermal stratification and natural circulation have been researched. It’s found that at a constant flow area when heat transfer tubes’ number increased outlet temperature of PRHR HX is lower, the whole water temperature of IRWST is higher, thermal stratification and natural circulation are more oblivious. At a constant mass flow when inlet temperature of PRHR HX increased, inlet flow velocity increases and outlet temperature is higher. But on the other hand the cooling rate increases at the same time, the average temperature of IRWST is higher, the range of thermal stratification expands and the velocity of natural circulation increases.

The initiation events analysis and evaluation were the beginning of nuclear safety analysis and probabilistic safety assessment, and it was the key points of the nuclear safety analysis. The main methods of the initiation events analysis are reference to existing lists and reports, operating experience, project evaluation and logic diagrams analysis. Currently, the initiation events analysis method and experiences both focused on pressurized water reactor but there are no general theories for Fluoride Salt-Cooled High Temperature Reactor (FHR). With FHR’s research and development, the initiation events analysis and evaluation was increasingly important. Based on the FHR’s design, the theories and methods of initiation events analysis would be researched and developed. From the FHR’s design, the systems, subsystems and components are divided to identify the safety functions of them. Base on the safety functions, the logical analysis and accident analysis calculation method would be combined to study FHR’s initiation events. The theory of analysis would be developed and the analysis method system would be discussed. Finally, the preliminary initiation events list of FHR will be discussed and researched. The results would help TMSR’s reactor designs and nuclear safety analysis.

The failure mode of Sn-4.0Ag-0.5Cu solder joints under the intermediate strain rates is explored. Tensile and shear tests were performed on miniature solder joints at two different displacement rates, i.e., 5 and 50 mm/s. For the uniaxial tensile test, brittle failure occurred at the intermetallic compound (IMC)/Cu pad interface on the board side under both loading rates. However, for shear loading at 5 mm/s, ductile failure along the solder matrix was detected. With the increase of the shear loading rate up to 50 mm/s, the failure mode is converted into the (ductile+brittle) mixed mode. Numerical simulation results based on the extracted elastic-plastic solder properties indicate that the maximum stress concentration is located at the interfaces of the board side, which indicates the weak position of the crack initiation. The plastic deformation of the solder alloy is significantly suppressed at 50 mm/s loading rate. The amplitude of the solder plastic deformation under shear loading is much greater than that for the tensile test. Furthermore, most plastic deformation under shear loading is in the solder alloy adjacent to solder/IMC interface of the board side and no visible plastic deformation is detected in the middle of solder ball, which results in cohesive failure at the rate of 5 mm/s and the (ductile+brittle) mixed failure mode at the rate of 50 mm/s.

The influence of the mass flow rate of secondary air on the gas/particle flow characteristics of a double swirl flow burner, in the near-burner region, was measured by a three-component particle-dynamics anemometer, in conjunction with a gas/particle two-phase test facility. Velocities, particle volume flux profiles, and normalized particle number concentrations were obtained. The relationship between the gas/particle flows and the combustion characteristics of the burners was discussed. For different mass flow rates of secondary air, annular recirculation zones formed only in the region of r/d=0.3–0.6 at x/d=0.1–0.3. With an increasing mass flow rate of secondary air, the peaks of the root mean square (RMS) axial fluctuating velocities, radial mean velocities, RMS radial fluctuating velocities, and tangential velocities all increased, while the recirculation increased slightly. There was a low particle volume flux in the central zone of the burner. At x/d=0.1–0.7, the profiles of particle volume flux had two peaks in the secondary air flow zone near the wall. With an increasing mass flow rate of secondary air, the peak of particle volume flux in the secondary air flow zone decreased, but the peak of particle volume flux near the wall increased. In section x/d=0.1–0.5, the particle diameter in the central zone of the burner was always less than the particle diameter at other locations.