Method for constructing and controlling turbulence form of fluid inside reactor pressure vessel

Abstract

A method for constructing and controlling a turbulence form of a fluid inside a reactor pressure vessel. The method comprises: step S1, obtaining related data of vortexes, a temporal distribution characteristic, a spatial distribution characteristic and a power spectral density of turbulence in an inlet area of each fuel assembly, and a turbulence energy spectrum; step S2, obtaining a water gap change situation of each fuel assembly, and on the basis of the water gap change situation of each fuel assembly, determining whether a three-dimensional flow field model of each fuel assembly needs to be corrected, until a water gap of each fuel assembly does not change; step S3, obtaining a reactor core nuclear power fluctuation amplitude value between fuel assemblies; and step S4, determining whether the reactor core nuclear power fluctuation amplitude value falls within a target range, and if the reactor core nuclear power fluctuation amplitude value does not fall within the target range, changing a three-dimensional structure inside a reactor pressure vessel until the reactor core nuclear power fluctuation amplitude value falls within the target range. The method for constructing and controlling a turbulence form of a fluid inside a reactor pressure vessel can eliminate or ameliorate the nuclear power fluctuation of a reactor core, reduce the wear of grids of the reactor core, and ensure the safety of reactor operation.

Description

Construction and control method of turbulent flow pattern inside reactor pressure vessel Technical Field

[0001] The present application relates to the field of nuclear power technology, and more specifically, to a method for constructing and controlling the turbulent flow morphology of fluid inside a reactor pressure vessel. Background Art

[0002] The main body of a pressurized water reactor (PWR) consists of the reactor pressure vessel (RPV), reactor internals (RVI), control rod drive mechanism, core components, core instrumentation, and related parts. The flow field within the reactor is shown in Figure 1. Coolant entering the lower chamber from the RPV inlet and annular downcomer generates a series of vortexes of varying sizes in the RPV's lower chamber due to the driving force of the reactor coolant main pump and the drastic changes in the RPV's internal flow channel geometry. This results in uneven flow distribution to fuel assemblies at different locations in the core and significant pulsation in the flow corresponding to each fuel assembly.

[0003] At present, it is difficult to optimize the flow channel geometry inside the reactor pressure vessel in a reasonable way, construct and control the turbulent flow pattern in the bottom chamber of the reactor pressure vessel, thereby improving the impact of the fluid on the vibration behavior of the fuel assembly and slowing down or eliminating the low-frequency oscillation phenomenon of the fuel assembly.

[0004] Summary of the Invention

[0005] The technical problem to be solved by the present application is to provide a method for constructing and controlling the turbulent flow morphology of the fluid inside the reactor pressure vessel.

[0006] The technical solution adopted by the embodiment of the present application to solve its technical problems is: constructing a method for constructing and controlling the turbulent morphology of the fluid inside the reactor pressure vessel, including: step S1, establishing a three-dimensional structural model of the interior of the reactor pressure vessel, and conducting an analysis or experiment of the flow field inside the reactor to obtain a dynamic flow field of the fluid inside the reactor, obtaining relevant data of vortices and the time distribution characteristics and spatial distribution characteristics of turbulence in the inlet area of ​​each fuel assembly based on the dynamic flow field, and obtaining the power spectrum density of turbulence in the inlet area of ​​each fuel assembly based on the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, performing energy spectrum analysis on the turbulence, and obtaining a turbulent energy spectrum; step S2, establishing a three-dimensional flow field model of the fuel assembly, and analyzing the turbulence based on the turbulence-related data and the turbulence in the inlet area of ​​each fuel assembly. The time distribution characteristics, spatial distribution characteristics and turbulence energy spectrum of the inlet area are analyzed through the low-frequency oscillation of the fuel assembly to obtain the change of the water gap of the fuel assembly, and whether the three-dimensional flow field model of the fuel assembly needs to be corrected according to the change of the water gap of the fuel assembly, until the water gap of the fuel assembly does not change; step S3, establish a core calculation model, and obtain the core nuclear power fluctuation amplitude value between each fuel assembly according to the low-frequency oscillation analysis result of the fuel assembly; step S4, judge whether the core nuclear power fluctuation amplitude value between each fuel assembly falls into the target interval, if not, change the internal three-dimensional structure of the reactor pressure vessel, and repeat steps S1 to S4 until the core nuclear power fluctuation amplitude value falls into the target interval.

[0007] In some embodiments, the step S1 includes: step S11, establishing a three-dimensional structural model of the interior of the reactor pressure vessel, conducting an analysis or experiment on the flow field inside the reactor, and obtaining a dynamic flow field of the fluid inside the reactor; step S12, obtaining relevant data of the vortex by image recognition based on the dynamic flow field; step S13, extracting the average flow rate of each fuel assembly inlet area in a representative time period through the dynamic flow field, and obtaining the spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; step S14, extracting the change value of the flow rate of each fuel assembly inlet area over time through the dynamic flow field, and obtaining the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; step S15, performing self-power spectrum analysis on the flow rate of each fuel assembly inlet area based on the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, and obtaining the power spectrum density of the turbulence in the inlet area of ​​each fuel assembly. step S16, performing energy spectrum analysis on the turbulence based on the scale of the vortex and the power spectrum density of the turbulence in the inlet area of ​​each fuel assembly, and obtaining the turbulence energy spectrum.

[0008] In some embodiments, the step S13 includes: extracting the flow average value of each fuel assembly inlet area in a representative time period through the dynamic flow field, obtaining the flow distribution coefficient of each fuel assembly inlet area based on the flow average value of each fuel assembly inlet area, and obtaining the maximum value, minimum value and overall spatial standard deviation of the flow distribution of each fuel assembly inlet area based on the flow distribution coefficient, thereby obtaining the spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; and / or, the step S14 includes: extracting the flow change value of each fuel assembly inlet area over time through the dynamic flow field, obtaining the fluid pulsation value of each fuel assembly inlet area based on the flow change value of each fuel assembly inlet area over time, and obtaining the standard deviation value of the flow change of each fuel assembly inlet area over time based on the fluid pulsation value, thereby obtaining the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly.

[0009] In some embodiments, in step S1 , the vortex-related data includes the number, size, frequency, and location of the vortices.

[0010] In some embodiments, the step S2 includes: step S21, establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation and oscillation position of the fuel assembly through low-frequency oscillation analysis of the fuel assembly based on the relevant data of the vortex, the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, and the spatial distribution characteristics; step S22, performing a fuel assembly water gap analysis based on the deformation and oscillation position of the fuel assembly to determine whether the fuel assembly water gap has changed; if the fuel assembly water gap has changed, adjusting the fuel assembly three-dimensional flow field model, and repeating the steps S21 and S22 until the fuel assembly water gap has not changed.

[0011] In some embodiments, the step S21 includes: step S211, establishing a three-dimensional flow field model with the core as the object, and obtaining the lateral flow distribution inside the core through fluid dynamics analysis based on the relevant data of the vortex, the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, and the spatial distribution characteristics; step S212, obtaining the lateral loads on different positions of the fuel assembly based on the lateral flow distribution inside the core; step S213, obtaining the flow-induced vibration of the fuel assembly based on the lateral loads on different positions of the fuel assembly; step S214, obtaining the deformation and oscillation position of the fuel assembly based on the flow-induced vibration of the fuel assembly.

[0012] In some embodiments, the step S21 includes: establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation and oscillation position of the fuel assembly through low-frequency oscillation analysis of the fuel assembly based on the relevant data of the vortex and the turbulence energy spectrum.

[0013] In some embodiments, the step S3 includes: step S31, establishing a core calculation model; step S32, obtaining the core neutron transport situation based on the change of the water gap of the fuel assembly; step S33, obtaining the heating power of the fuel assembly and the partition loading situation of the fuel assembly; step S34, obtaining the core nuclear power fluctuation amplitude value between each fuel assembly based on the heating power of the fuel assembly, the core neutron transport situation, the low-frequency oscillation analysis results of the fuel assembly and the partition loading situation of the fuel assembly.

[0014] In some embodiments, step S34 includes: according to the heating power of the fuel assembly, the neutron transport conditions of the core, the low-frequency oscillation analysis results of the fuel assembly and the partitioned loading conditions of the fuel assembly, the core nuclear power fluctuation analysis of the coupled nuclear-thermal-mechanical load is performed to obtain the core nuclear power fluctuation amplitude value between each fuel assembly.

[0015] In some embodiments, the step S2 includes: step S21, establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation and oscillation position of the fuel assembly through low-frequency oscillation analysis of the fuel assembly based on the relevant data of the vortex, the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, and the spatial distribution characteristics; step S22, performing a fuel assembly water gap analysis based on the deformation and oscillation position of the fuel assembly to determine whether the fuel assembly water gap has changed; if the fuel assembly water gap changes, adjusting the fuel assembly three-dimensional flow field model, and repeating the steps S21 and S22 until the fuel assembly water gap does not change; the step S34 includes: obtaining the core nuclear power fluctuation amplitude value between each fuel assembly based on the heating power of the fuel assembly, the core neutron transport, the oscillation position of the fuel assembly, and the partitioned loading of the fuel assembly.

[0016] In some embodiments, the step S4 includes: step S41, judging whether the core nuclear power fluctuation amplitude value between each fuel assembly falls within the target range; if not, establishing the target limit range of the time distribution characteristics and spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly based on the target range; step S42, judging whether the time distribution characteristics and spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly obtained in the step S1 fall within the respective target limit ranges; if not, changing the internal three-dimensional structure of the reactor pressure vessel, and repeating the step S1 until the time distribution characteristics and spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly obtained in the step S1 fall within the respective target limit ranges; step S43, repeating the steps S2 and S3, judging whether the core nuclear power fluctuation amplitude value measured in the step S3 this time falls within the target range; if not, re-changing the internal three-dimensional structure of the reactor pressure vessel, and repeating the steps S1, S42 and S43 until the core nuclear power fluctuation amplitude value is within the target range.

[0017] In some embodiments, the step S1 further includes: step S17, obtaining a flow correlation coefficient between the inlet area of ​​each fuel assembly and the inlet areas of other fuel assemblies based on the dynamic flow field and the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly;

[0018] The step S2 includes: step S21, establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation, oscillation position, and low-frequency oscillation frequency of the fuel assembly through low-frequency oscillation analysis of the fuel assembly based on the relevant data of the vortex and the time distribution characteristics and spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; step S22, performing a fuel assembly water gap analysis based on the deformation and oscillation position of the fuel assembly to determine whether the fuel assembly water gap has changed; if the fuel assembly water gap has changed, adjusting the fuel assembly three-dimensional flow field model, and repeating steps S21 and S22 until the fuel assembly water gap has not changed;

[0019] The step S4 includes: step S411, judging whether the core nuclear power fluctuation amplitude values ​​between each fuel assembly fall within the target interval; if not, performing multi-parameter fitting on the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, the turbulence energy spectrum, the low-frequency oscillation frequency, and the core nuclear power fluctuation amplitude value to obtain a multi-parameter fitting relationship; step S412, according to the multi-parameter fitting relationship, based on the target interval of the nuclear power fluctuation amplitude value, establishing a target limit range of the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum that meets the target interval condition; step S42, judging the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in step S1. The value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum fall within their respective target limit ranges; if not, the internal three-dimensional structure of the reactor pressure vessel is changed, and step S1 is repeated until the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in step S1 all fall within their respective target limit ranges; step S43, repeating steps S2 and S3, and judging whether the core nuclear power fluctuation amplitude value measured in step S3 this time falls within the target interval; if not, the internal three-dimensional structure of the reactor pressure vessel is changed again, and steps S1, S42, and S43 are repeated until the core nuclear power fluctuation amplitude value falls within the target interval.

[0020] In some embodiments, the step S42 includes: determining whether the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in the step S1 fall within their respective target limit ranges; if not, establishing at least one turbulence control region in the reactor pressure vessel, improving the dynamic flow field morphology in the reactor pressure vessel by adding a vortex breaking-up component in the at least one turbulence control region, and repeating the step S1 until the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in the step S1 all fall within their respective target limit ranges;

[0021] The step S43 includes: repeating the step S2 and the step S3, and judging whether the core nuclear power fluctuation amplitude value measured in the step S3 falls within the target range; if not, improving the dynamic flow field morphology in the reactor pressure vessel again, and repeating the step S1, the step S42 and the step S43 until the core nuclear power fluctuation amplitude value falls within the target range.

[0022] In some embodiments, the number of the at least one turbulence control region is three, namely, an annular descending section region, a lower chamber region, and an upper chamber region.

[0023] In some embodiments, in step S17, the flow correlation coefficient is calculated using Formula 1, and the calculation process of Formula 1 is:

[0024] Among them, C ab is the flow correlation coefficient between the inlet area of ​​fuel assembly a and the inlet area of ​​fuel assembly b; is the average value of the fluid pulsation values ​​of all fuel assembly inlet areas centered on the inlet area of ​​fuel assembly a; is the average value of the fluid pulsation values ​​of all fuel assembly inlet areas centered on the inlet area of ​​fuel assembly b; is the variance of the fluid pulsation values ​​at the inlet area of ​​all fuel assemblies centered at the inlet area of ​​fuel assembly a; is the variance of the fluid pulsation values ​​at the inlet area of ​​all fuel assemblies centered at the inlet area of ​​fuel assembly b.

[0025] The implementation of the embodiments of the present application has at least the following beneficial effects: the present application constructs the turbulent morphology inside the reactor pressure vessel by comprehensively considering the temporal and spatial distribution characteristics of turbulence inside the reactor, the power spectrum density, the turbulent energy spectrum, and the relevant data of the vortices formed, through construction step S1, and provides basic data support for the control of the turbulent morphology; the present application optimizes the flow field inside the reactor, improves the flow field conditions in the core inlet area, reduces the coherence of the flow between the fuel assembly inlet areas, and avoids changes in the water gap of the fuel assembly during turbulence, which leads to problems such as low-frequency oscillation of the fuel assembly; the present application also corrects the three-dimensional structure inside the reactor pressure vessel through construction steps S3 and S4, thereby further improving the three-dimensional flow field inside the reactor and controlling the turbulent morphology, thereby ensuring that the water gap of the fuel assembly does not change while further reducing the flow coherence in the fuel assembly inlet area, so that the core nuclear power fluctuation amplitude value is within the target range, thereby eliminating or improving the problems of core nuclear power fluctuation and fuel assembly low-frequency oscillation; and ultimately achieving the comprehensive result of reducing core frame wear and ensuring reactor operation safety. BRIEF DESCRIPTION OF THE DRAWINGS

[0026] The present application will be further described below with reference to the accompanying drawings and embodiments, in which:

[0027] FIG1 is a schematic diagram of the internal flow field of a reactor pressure vessel in the prior art;

[0028] FIG2 is a flow chart of a method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel in one embodiment of the present application;

[0029] FIG3 is a flowchart of step S1 in the flowchart shown in FIG2 ;

[0030] FIG4 is a flowchart of step S2 in the flowchart shown in FIG2 ;

[0031] FIG5 is a flowchart of step S21 in the flowchart shown in FIG4 ;

[0032] FIG6 is a flowchart of step S3 in the flowchart shown in FIG2 ;

[0033] FIG7 is a flowchart of step S4 in the flowchart shown in FIG2 ;

[0034] FIG8 is a flowchart of step S4 in the flowchart shown in FIG2 in another embodiment. DETAILED DESCRIPTION

[0035] In order to have a clearer understanding of the technical features, purposes and effects of the present application, the specific embodiments of the present application are now described in detail with reference to the accompanying drawings. In the following description, it should be understood that the directions or positional relationships indicated by "front", "back", "up", "down", "left", "right", "longitudinal", "horizontal", "vertical", "horizontal", "top", "bottom", "inside", "outside", "head", "tail", etc. are based on the directions or positional relationships shown in the accompanying drawings and are constructed and operated in specific directions. They are only for the convenience of describing the technical solution and do not indicate that the devices or components referred to must have specific directions. Therefore, they should not be understood as limitations on the present application.

[0036] In the following description, specific details such as specific system structures and techniques are provided for purposes of illustration rather than limitation to facilitate a thorough understanding of the embodiments of the present application. However, it will be apparent to those skilled in the art that the present application may be implemented in other embodiments without these specific details. In other cases, detailed descriptions of well-known systems, devices, circuits, and methods are omitted to avoid obscuring the description of the present application with unnecessary detail.

[0037] As shown in FIG2 , the present application constructs a method for constructing and controlling the turbulent flow morphology of fluid inside a reactor pressure vessel, including:

[0038] Step S1: Establish a three-dimensional structural model of the reactor pressure vessel interior and conduct an analysis or experiment of the reactor's internal flow field to obtain the dynamic flow field of the reactor's internal fluid. Based on the dynamic flow field, obtain relevant data on vortices and the temporal and spatial distribution characteristics of turbulence at the inlet area of ​​each fuel assembly. Based on the temporal distribution characteristics of turbulence at the inlet area of ​​each fuel assembly, obtain the power spectrum density of turbulence at the inlet area of ​​each fuel assembly, perform energy spectrum analysis on the turbulence, and obtain the turbulence energy spectrum. Step S2: Establish a three-dimensional flow field model of the fuel assembly. Based on the turbulence-related data, the temporal and spatial distribution characteristics of turbulence at the inlet area of ​​each fuel assembly, and the turbulence energy spectrum, analyze the low-frequency oscillation of the fuel assembly to obtain the change in the fuel assembly water gap. Based on the change in the fuel assembly water gap, determine whether the three-dimensional flow field model of the fuel assembly needs to be modified until the fuel assembly water gap does not change. Step S3: Establish a core calculation model. Based on the results of the fuel assembly low-frequency oscillation analysis, obtain the core nuclear power fluctuation amplitude value between each fuel assembly. Step S4: Determine whether the core nuclear power fluctuation amplitude values ​​between the fuel assemblies fall within the target range. If not, change the internal three-dimensional structure of the reactor pressure vessel and repeat steps S1 to S4 until the core nuclear power fluctuation amplitude values ​​fall within the target range.

[0039] It should be understood that in step S4, the "target range" is a known range of values ​​for the reactor type of the nuclear power plant. Different reactor types have slightly different requirements for the target range. In this embodiment, the target range is 5% to 10% (inclusive).

[0040] This application completes the construction of the turbulence morphology by obtaining the relevant data of the vortex in step S1, the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, the spatial distribution characteristics, the power spectrum density, and the turbulence energy spectrum, and provides data support for subsequent analysis and control of the turbulence morphology. This application improves the flow field inside the core by taking into account the time characteristics of the fluid and the relevant data of the vortex generated, and combining them with the spatial distribution characteristics of the fluid to ensure that the fluid entering the core is in a relatively stable and uniform state, improve the turbulence morphology, and thus eliminate or improve the problem of low-frequency oscillation of the fuel assembly caused by the large coherence between the fuel assemblies, the pulsating response of the flow rate, and the momentum transport of the vortex generated by the turbulence, so as to control it within the required range. This application analyzes the core nuclear power fluctuations so that the turbulence morphology within the core meets the reactor's requirements for the core nuclear power fluctuation amplitude under the premise of meeting the conditions established in step S2. Step S4 modifies the three-dimensional structure inside the reactor pressure vessel, thereby further improving the three-dimensional flow field inside the reactor and controlling the turbulent morphology. This ensures that the fuel assembly water gap remains unchanged while further reducing the flow coherence in the fuel assembly inlet area, thereby ensuring that the core nuclear power fluctuation amplitude is within the target range, thereby eliminating or improving the core nuclear power fluctuation problem. Furthermore, by eliminating or improving core nuclear power fluctuations, low-frequency oscillations of fuel assemblies, and other phenomena, the effects of reducing core grid wear and ensuring reactor operation safety are achieved.

[0041] As shown in FIG3 , step S1 includes the following steps: Step S11: Establishing a three-dimensional structural model of the reactor pressure vessel interior, conducting an analysis or experiment on the reactor's internal flow field, and obtaining the dynamic flow field of the fluid within the reactor. Step S12: Obtaining vortex-related data through image recognition based on the dynamic flow field. Step S13: Extracting the average flow rate of each fuel assembly inlet region during a representative time period through the dynamic flow field to obtain the spatial distribution characteristics of turbulence at each fuel assembly inlet region. Step S14: Extracting the time-varying values ​​of the flow rate at each fuel assembly inlet region through the dynamic flow field to obtain the temporal distribution characteristics of turbulence at each fuel assembly inlet region. Step S15: Based on the temporal distribution characteristics of turbulence at each fuel assembly inlet region, performing autopower spectrum analysis on the flow rate at each fuel assembly inlet region to obtain the power spectrum density of turbulence at each fuel assembly inlet region. Step S16: Performing energy spectrum analysis on the turbulence based on the scale of the vortex and the power spectrum density of turbulence at each fuel assembly inlet region to obtain the turbulence energy spectrum. Step S17: Obtain the flow correlation coefficient between each fuel assembly inlet area and the remaining fuel assembly inlet areas based on the dynamic flow field and the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly.

[0042] In some embodiments, in step S11, "reactor internal flow field analysis" can be performed using computational fluid dynamics (CFO) software (representative software includes ANSYS CFX, ANSYS FLUENT, STAR CCM+, etc.). The software establishes a three-dimensional flow field model within the reactor and divides the network. By setting boundary conditions and turbulence models, and using its post-processing functions, the dynamic flow field conditions at various locations or cross-sections are obtained. The "experiment" can be performed by creating a visual scaled model (typically ranging from 1:5 to 1:10). This model primarily includes the reactor pressure vessel, internal components, core dummy components, and inlet and outlet pipes. The inlet and outlet pipes are connected to a test loop system to simulate the coolant circulation process within the reactor. During this process, pressure sensors can be placed at various typical locations in the test apparatus (such as the inlet and outlet pipes, downcomer, lower chamber, and core inlet and outlet) to measure the pressure drop between sections. Flow rate data at each fuel assembly inlet can also be measured by installing a vortex flowmeter at each fuel assembly inlet. The experimental device can also be made transparent by adding tracer particles to the coolant and using particle image velocimetry (PIV) to take photos to obtain the dynamic flow field conditions in typical areas (such as the downcomer, lower chamber, and core inlet).

[0043] In some embodiments, in step S12, the "image recognition method" can be performed using image processing software, such as extracting velocity distribution (scalar or vector), pressure distribution, vortex representative parameters (such as turbulent kinetic energy, turbulent dissipation rate, etc.) or flow data (such as the inlet area of ​​each fuel assembly) of certain typical cross-sections or inlets through the post-processing function of computational fluid dynamics (CFO) software, and obtaining vortex-related data by analyzing velocity vector diagrams at different positions and times. Alternatively, vortex-related data can be obtained by processing image analysis software (such as MATLAB software). Alternatively, manual recognition can be performed to manually measure vortex-related data based on the dynamic flow field obtained.

[0044] In some embodiments, in step S12, the vortex-related data includes the number, scale, frequency, and location of the vortex.

[0045] In some embodiments, step S13 includes: extracting the average flow rate of each fuel assembly inlet area in a representative time period through a dynamic flow field, obtaining the flow distribution coefficient of each fuel assembly inlet area based on the average flow rate of each fuel assembly inlet area, and obtaining the maximum value, minimum value and overall spatial standard deviation of the flow distribution of each fuel assembly inlet area based on the flow distribution coefficient, thereby obtaining the spatial distribution characteristics of turbulence in the inlet area of ​​each fuel assembly.

[0046] The representative time period is 5-50 seconds.

[0047] It should be understood that the "flow distribution coefficient" is the average flow rate at the inlet of a fuel assembly during a representative time period.

[0048] In some embodiments, step S14 includes: extracting the time-varying flow rate values ​​of the inlet areas of each fuel assembly through a dynamic flow field, obtaining the fluid pulsation values ​​of the inlet areas of each fuel assembly based on the time-varying flow rate values ​​of the inlet areas of each fuel assembly, and obtaining the standard deviation values ​​of the flow rate of the inlet areas of each fuel assembly based on the fluid pulsation values, thereby obtaining the time distribution characteristics of the turbulence in the inlet areas of each fuel assembly.

[0049] It should be understood that both step S13 and step S14 can be obtained by the aforementioned flow field analysis software and computational fluid dynamics (CFO) software. Step S13 and step S14 can be performed in different orders or simultaneously.

[0050] In step S15, an autopower spectrum analysis is performed on the flow rate at each fuel assembly inlet. This can be achieved using the Fourier transform method, using the temporal distribution characteristics of the turbulent flow at each fuel assembly inlet as input. This step S15 uses autopower spectrum analysis to convert the temporal distribution characteristics of the flow rate at each fuel assembly inlet into frequency domain distribution characteristics. This converts time domain data into frequency domain data, thereby obtaining the pulsation frequency of the turbulent flow at each fuel assembly inlet.

[0051] In some embodiments, the energy spectrum analysis of the turbulence (the turbulence energy spectrum is a function that expresses the contribution of turbulent velocity fluctuations of different frequencies to the turbulent kinetic energy) can be performed based on Kolmogorov Theory (that is, the energy of large eddies is transferred to the energy of small eddies, and the small eddies dissipate the energy at a smaller scale. Because the stable state is considered, the input energy, the transferred energy, and the final dissipated energy are equal) to calculate the turbulence energy spectrum (the kinetic energy and frequency of the turbulence), and then obtain the relationship between the turbulence scale, the turbulence pulsation frequency, and the turbulence energy spectrum.

[0052] It should be understood that the energy spectrum analysis of the turbulence can be achieved through programming software such as MATLAB.

[0053] In some embodiments, in step S17, the “flow correlation coefficient” can be calculated using Formula 1, and the calculation process of Formula 1 is:

[0054] Among them, C ab is the flow correlation coefficient between the inlet area of ​​fuel assembly a and the inlet area of ​​fuel assembly b; is the average value of the fluid pulsation value of all fuel assembly inlet areas centered on the inlet area of ​​fuel assembly a is the average value of the fluid pulsation value of all fuel assembly inlet areas centered on the inlet area of ​​fuel assembly b is the variance of the fluid pulsation values ​​at the inlet area of ​​all fuel assemblies centered at the inlet area of ​​fuel assembly a; is the variance of the fluid pulsation values ​​at the inlet area of ​​all fuel assemblies centered at the inlet area of ​​fuel assembly b.

[0055] For example, when the reactor core includes 3*3 fuel assemblies, the number of fuel assemblies in Formula 1 is It is the average value of the fluid pulsation values ​​of the inlet areas of nine fuel assemblies within a 3*3 range, with the inlet area of ​​fuel assembly a (for example, one of the 3*3 fuel assemblies is fuel assembly a) as the center. is the variance of the fluid pulsation values ​​at the inlet areas of nine fuel assemblies within a 3*3 range centered at the inlet area of ​​fuel assembly a. and Same thing.

[0056] Through step S17, the correlation between the fuel assemblies within the core can be obtained. For example, when the flow rate at a certain location increases, does the flow rate at other locations increase (positive correlation), decrease (negative correlation), or have no significant effect? ​​By incorporating the correlation of the fuel assembly into the analysis, the correlation between the fuel assemblies can be controlled within a certain range through subsequent steps, thereby reducing the impact of turbulence on the fuel assemblies.

[0057] It should be understood that step S17 can be located before step S15 (or step S16) or performed simultaneously therewith in step S17, and it only needs to be placed after step S14.

[0058] As shown in Figure 4, step S2 includes the following: Step S21: Establishing a three-dimensional flow field model based on the core. Based on relevant data on vortices and the temporal and spatial distribution characteristics of turbulence at the inlet areas of each fuel assembly, low-frequency oscillation analysis of the fuel assembly is performed to determine the deformation and oscillation position of the fuel assembly. Step S22: Based on the deformation and oscillation position of the fuel assembly, a water gap analysis of the fuel assembly is performed to determine whether the water gap has changed. If the water gap of the fuel assembly has changed, the three-dimensional flow field model of the fuel assembly is adjusted, and steps S21 and S22 are repeated until the water gap of the fuel assembly remains unchanged.

[0059] Specifically, as shown in Figure 5, step S21 includes the following: Step S211: Establishing a three-dimensional flow field model with the core as the object, and obtaining the lateral flow distribution within the core through fluid dynamics analysis based on the relevant data of vortices, the temporal distribution characteristics of turbulence at the inlet area of ​​each fuel assembly, and the spatial distribution characteristics. Step S212: Obtaining the lateral loads on different positions of the fuel assembly based on the lateral flow distribution within the core. Step S213: Obtaining the flow-induced vibration of the fuel assembly based on the lateral loads on different positions of the fuel assembly. Step S214: Obtaining the deformation and oscillation position of the fuel assembly based on the flow-induced vibration of the fuel assembly.

[0060] In some embodiments, step S211 can be implemented by computational fluid dynamics (CFO) software (representative software includes ANSYS CFX, ANSYS FLUENT, STAR CCM+, etc.).

[0061] Specifically, in the process of establishing a three-dimensional flow field model, it can also be achieved through the computational fluid dynamics (CFO) software, wherein the fuel assembly can be set as a porous medium by assembling it as a unit (the simulation structure of the resistance coefficient and porosity at each position inside the porous medium can be obtained through existing experiments or analytical calculations). Furthermore, the computational fluid dynamics (CFO) software can be used to calculate the flow field conditions inside the core by taking the time distribution characteristics of turbulence at the inlet area of ​​each fuel assembly, the number, frequency, and scale of vortices as input. Then, through the post-processing function of the computational fluid dynamics (CFO) software, the pressure, velocity and other parameter data at each position inside the core are obtained, that is, the fluid response characteristics of the core area are obtained, and then the cross-flow distribution conditions inside the core are obtained. The mapping relationship between the momentum transport of turbulence, the pressure dynamic response and the fluid characteristic response of the core area can be analyzed, providing input for the low-frequency oscillation analysis of the fuel assembly.

[0062] It should be understood that the "turbulent momentum transport" refers to the temporal variation of the mass flow rate and velocity direction of the turbulent flow. The "pressure dynamic response" refers to the temporal variation of the pressure at each fuel assembly inlet region. Both the turbulent momentum transport and pressure dynamic response can be obtained from the "temporal distribution characteristics of turbulence at each fuel assembly inlet region" in step S14. The "core region fluid characteristic response" refers to the flow rate, velocity, and pressure distribution within the fuel assembly. Because the porous media representing the fuel assemblies in the constructed three-dimensional flow field model have gaps (i.e., gaps between each fuel assembly), transverse flow occurs in these gaps (the mainstream coolant flows upward from the core inlet (at the bottom) and out of the core outlet. However, due to the complex structure of the fuel assemblies and the uneven inlet flow rate and pressure, some transverse flow occurs between the fuel assemblies, which is referred to as transverse flow). The obtained core region fluid characteristic response refers to the flow rate, velocity, and pressure distribution within the fuel assembly. This flow rate includes both transverse and longitudinal flow. The transverse flow distribution within the core can be obtained from the core region fluid response characteristics.

[0063] It is important to understand that existing in-pile fuel assembly lateral flow and fuel deformation parameters are calculated based on the "time-averaged" flow rate over a period of time and cannot reflect the actual lateral flow state and deformation phenomena at different times. This application, through a comprehensive analysis of the temporal and spatial distribution characteristics of the three-dimensional flow field, can not only reflect the lateral flow and fuel deformation of the in-pile fuel assembly at a certain time or time period, but also reflect the lateral flow and fuel deformation of the in-pile fuel assembly at different times, making the analysis of the in-pile lateral flow more comprehensive and ensuring the accuracy of the analysis in subsequent steps.

[0064] In other optional embodiments, since the turbulence energy spectrum can reflect the distribution of turbulent kinetic energy among different large and small eddies through fluctuations, and thus intuitively provides the basic features of the turbulent field from one side, step S21 can also be implemented in some embodiments by establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation and oscillation position of the fuel assembly through low-frequency oscillation analysis of the fuel assembly based on the relevant data of the eddies and the turbulence energy spectrum.

[0065] In some embodiments, in step S212 and step S213, fluid-structure coupling software (such as ANSYS or other dedicated software) can be used to calculate the lateral loads on the fuel assembly at different positions, and then the flow-induced vibration of the fuel assembly can be calculated based on the lateral loads using the fluid-structure coupling software.

[0066] By analyzing this flow-induced vibration, we can determine whether the vibration of the fuel assembly is within the required limit. When it is within the required limit, the fuel assembly's flow-elastic instability rate (i.e., the ratio of the critical cross-flow velocity to the effective excitation fluid velocity) is less than the limit, and flow-elastic instability will not occur. The maximum vibration amplitude of the fuel rod caused by turbulence under the influence of the cross-flow will not exceed the limit. Over the designed lifespan, the fuel rod will not experience excessive wear due to flow-induced vibration. This flow-induced vibration analysis further ensures the safety of the fuel assembly during operation.

[0067] In some embodiments, in step S214, conclusions regarding the deformation, oscillation position, and low-frequency oscillation frequency of the fuel assembly can be obtained by calculating core neutron heating. It should be understood that core neutron heating is calculated using dedicated nuclear power core thermal calculation software based on the fuel assembly's enrichment, moderator concentration, neutron count, and other factors.

[0068] It should be understood that in step S22, the "water gap" refers to the gap between the fuel assemblies. If the fuel assembly deforms in step S24, causing a change in the oscillation position (i.e., the fuel rods deviate from their original positions due to oscillation), this will cause the water gap to change. "Adjusting the three-dimensional flow field model of the fuel assembly" refers to adjusting the water gap between the fuel assemblies.

[0069] As shown in FIG6 , step S3 includes: step S31, establishing a core calculation model; step S32, obtaining the core neutron transport situation based on the change of the water gap of the fuel assembly; step S33, obtaining the heating power of the fuel assembly and the partition loading situation of the fuel assembly; step S34, obtaining the core nuclear power fluctuation amplitude value between each fuel assembly based on the heating power of the fuel assembly, the core neutron transport situation, the low-frequency oscillation analysis results of the fuel assembly and the partition loading situation of the fuel assembly.

[0070] Among them, the "core calculation model" refers to the three-dimensional calculation model of the core shape (including each group of fuel assemblies) constructed according to the size, number of groups, and arrangement of the core fuel assemblies (the model of each group of internal fuel rods in this model can be appropriately simplified, for example, by replacing it with a porous medium model).

[0071] It's important to understand that "zoned fuel assembly loading" refers to the distribution of fuel elements within the core. For example, some large nuclear power plants typically employ a zoned loading scheme with non-uniform enrichment. For example, in the traditional three-zone loading method, fuel of varying enrichment levels is arranged radially along the core in three zones. Fuel elements with the highest enrichment are placed in the outermost zone, while fuel elements with lower enrichments are placed in the inner zone. This is done to flatten the radial power distribution and avoid high power peaks in the center. During refueling, the oldest assemblies are removed, other assemblies are moved inward, and new fuel assemblies are placed on the periphery. Some nuclear power plants also employ a low-neutron leakage loading method to improve fuel utilization, enhance plant availability and overall economic efficiency, and reduce leakage and neutron radiation exposure to the pressure vessel. This involves placing new fuel assemblies near the center of the core, while spent fuel is placed on the periphery. The "heat generation power of the fuel assembly" can be calculated based on parameters such as the heat release rate per unit volume of nuclear fuel and the dimensions of the fuel assemblies.

[0072] It should be understood that since the acquisition of the heating power of the fuel assembly and the partitioned loading status of the fuel assembly is not affected by the other steps of this method, there is no order relationship between step S33 and step S31 and step S32. Step S33 can be performed before step S31 and / or step S32, or it can be performed simultaneously with step S31 or step S32.

[0073] In some embodiments, the core nuclear power fluctuation amplitude values ​​between the fuel assemblies can be obtained through a core nuclear power fluctuation analysis coupled with nuclear, thermal, and mechanical loads. Specifically, step S34 includes obtaining the core nuclear power fluctuation amplitude values ​​between the fuel assemblies through a core nuclear power fluctuation analysis coupled with nuclear, thermal, and mechanical loads based on the fuel assembly heating power, core neutron transport conditions, fuel assembly low-frequency oscillation analysis results, and fuel assembly zoning loading conditions.

[0074] In some embodiments, the fuel assembly low-frequency oscillation analysis results used in step S34 are based on the oscillation position of the fuel assembly. That is, step S34 further includes obtaining the core nuclear power fluctuation amplitude values ​​between the fuel assemblies based on the fuel assembly heating power, the core neutron transport conditions, the oscillation position of the fuel assembly, and the partitioned loading conditions of the fuel assembly.

[0075] It should be understood that the core neutron transport condition is related to the size of the gap between the fuel assemblies, and the flow-induced vibration will cause changes in the size of the gap between the fuel assemblies. Therefore, in step S32, the core neutron transport condition can be obtained through the flow-induced vibration condition obtained in step S23 and the changes in the water gap (gap) of the fuel assembly caused by it.

[0076] In some embodiments, the core neutron transport condition of step S32 can be obtained by calculating the core neutron heating according to the flow-induced vibration of the fuel assembly using the core thermal calculation software involved in step S24.

[0077] In some embodiments, in step S34, the "core power fluctuation analysis of coupled nuclear-thermo-mechanical loads" can be performed using nuclear-thermo-mechanical coupled calculation software specifically designed for nuclear power plants. Fuel assembly heating power, core neutron transport, and fuel assembly oscillation position are input as boundary conditions to obtain core power fluctuation amplitudes between fuel assemblies.

[0078] It should be understood that the "core nuclear power fluctuation amplitude value" is the fluctuation of the ratio of the power value at each area of ​​the core to the average power.

[0079] As shown in Figure 7, step S4 includes the following: Step S41: Determine whether the core nuclear power fluctuation amplitude values ​​between each fuel assembly fall within the target range. If not, establish target limit ranges for the temporal and spatial distribution characteristics of turbulence at the inlet region of each fuel assembly based on the target range. Step S42: Determine whether the temporal and spatial distribution characteristics of turbulence at the inlet region of each fuel assembly obtained in step S1 fall within their respective target limit ranges. If not, modify the three-dimensional internal structure of the reactor pressure vessel and repeat step S1 until the temporal and spatial distribution characteristics of turbulence at the inlet region of each fuel assembly obtained in step S1 fall within their respective target limit ranges. Step S43: Repeat steps S2 and S3 to determine whether the core nuclear power fluctuation amplitude values ​​measured in step S3 fall within the target range. If not, modify the three-dimensional internal structure of the reactor pressure vessel and repeat steps S1, S42, and S43 until the core nuclear power fluctuation amplitude values ​​are within the target range.

[0080] As shown in FIG8 , in some embodiments, step S41 includes: Step S411: Determining whether the core nuclear power fluctuation amplitude values ​​between each fuel assembly fall within a target range; if not, performing multi-parameter fitting on the flow distribution coefficient, fluid pulsation value, power spectral density, flow correlation coefficient, and turbulence energy spectrum, along with the low-frequency oscillation frequency and the core nuclear power fluctuation amplitude value, to obtain a multi-parameter fitting relationship; Step S412: Based on the multi-parameter fitting relationship and the target range of the nuclear power fluctuation amplitude value, establishing target limit ranges for the flow distribution coefficient, fluid pulsation value, power spectral density, flow correlation coefficient, and turbulence energy spectrum that meet the target range conditions.

[0081] Step S42 includes: determining whether the flow distribution coefficient, fluid pulsation value, power spectrum density, flow correlation coefficient, and turbulence energy spectrum measured in step S1 fall within their respective target limit ranges; if not, changing the internal three-dimensional structure of the reactor pressure vessel, and repeating step S1 until the flow distribution coefficient, fluid pulsation value, power spectrum density, flow correlation coefficient, and turbulence energy spectrum measured in step S1 all fall within their respective target limit ranges.

[0082] In some embodiments, step S411 can be performed by performing multi-parameter fitting using mathematical software (such as MATLAB) to obtain a multi-parameter fitting relationship. It should be understood that the "multi-parameter fitting relationship" can be a relational expression or a relational table. Step S412 can also be obtained by calculation using mathematical software (such as MATLAB).

[0083] In some embodiments, "changing the three-dimensional structure inside the reactor pressure vessel" can be achieved by adding / reducing / changing vortex breaking-up components inside the reactor pressure vessel. Therefore, step S42 includes: judging whether the flow distribution coefficient, fluid pulsation value, power spectrum density, flow correlation coefficient, and turbulence energy spectrum measured in step S1 fall within their respective target limit ranges. If not, at least one turbulence control area is established in the reactor pressure vessel, and the dynamic flow field morphology inside the reactor pressure vessel is improved by adding vortex breaking-up components in at least one turbulence control area, and step S1 is repeated until the flow distribution coefficient, fluid pulsation value, power spectrum density, flow correlation coefficient, and turbulence energy spectrum measured in step S1 all fall within their respective target limit ranges.

[0084] Step S43 includes repeating steps S2 and S3 to determine whether the core nuclear power fluctuation amplitude value measured in step S3 falls within the target range. If not, the dynamic flow field configuration within the reactor pressure vessel is further improved, and steps S1, S42, and S43 are repeated until the core nuclear power fluctuation amplitude value falls within the target range.

[0085] As shown in FIG1 , in this embodiment, there are three turbulence control regions, namely the annular descending section region, the lower chamber region, and the upper chamber region of the reactor pressure vessel.

[0086] In some other optional embodiments, vortex breaking components may be selectively added to the annular descending section area and / or the lower chamber area and / or the upper chamber area.

[0087] In some embodiments, the vortex breaking up component can be a mesh plate, grid plate, guide plate or other structure of various shapes, or other components that can break up large vortices into small vortices, so as to reduce the impact of the vortex on the reactor fuel assembly and improve the turbulent flow pattern in the reactor pressure vessel.

[0088] In some other optional embodiments, instead of adding vortex breaking components, the internal turbulence pattern can be improved by improving the original component structure in the reactor pressure vessel, such as changing its position.

[0089] It should be understood that in step S43, the purpose of "re-changing the three-dimensional structure inside the reactor pressure vessel" is to improve the dynamic flow field morphology inside the reactor pressure vessel, which can be achieved by changing the position of the original vortex breakup components, adding or reducing vortex breakup components, etc.

[0090] Since the flow correlation coefficient is the basis for judging the coherence of the core inlet, the correlation between the changing trends of the flow rates at various locations at the core inlet can be analyzed based on the correlation coefficient, which affects the subsequent nuclear power fluctuation amplitude value. Therefore, in some embodiments, in step S42, after improving the dynamic flow field morphology in the reactor pressure vessel and re-performing step S1, the flow correlation coefficient can be judged in advance after completing the measurement of the flow correlation coefficient and before proceeding to other steps, and then the nuclear power fluctuation amplitude can be roughly judged. If the flow correlation coefficient is obviously unreasonable, the subsequent measurement can be stopped when repeating to step S17 (this is because step S17 can be advanced to before steps S15 and S16), and the dynamic flow field morphology in the reactor pressure vessel can be directly improved again to save some workload.

[0091] Because different turbulent flow patterns can generate vortices of varying sizes and intensities, these vortices can lead to uneven flow distribution at the core inlet of different fuel assemblies, fluctuating at varying frequencies over time. The flow rates between individual fuel assemblies can also affect each other. The momentum and energy generated by these flows are transferred to the core, causing pressure and flow variations within the core. Transverse flows that cause large power fluctuations can also induce significant transverse hydraulic loads, leading to flow-induced vibrations, assembly oscillations, and grid wear.

[0092] It should be understood that under current technology, the flow field analysis of the lower chamber of the reactor and the flow field, structure and power analysis of the core fuel assembly are based on the "time-averaged" flow rate within a certain period of time. The relationship between the turbulence morphology, turbulence evolution, turbulence energy spectrum of the lower chamber of the reactor and the low-frequency oscillation of the fuel assembly and the core nuclear power fluctuation has not been established. There is no control technology for the turbulence morphology, making it difficult to eliminate or improve the impact on the core.

[0093] Through steps S411 and S412, target limits for various parameters are established, thereby providing result standards for improving the dynamic flow field morphology within the reactor pressure vessel. Furthermore, by constructing a vortex breakup component within the reactor turbulence control region, large vortices caused by turbulence are broken up into small vortices, changing the flow field within the pressure vessel, and thereby changing parameters such as the scale, frequency, and intensity of the turbulence, improving the flow field in the lower chamber and the flow field state of the core fuel assembly (including: reducing the difference between the maximum and minimum values ​​of the flow distribution coefficient, reducing the fluid pulsation value (standard deviation) of each fuel assembly, reducing the correlation between each fuel assembly, cross flow, fuel deformation, etc.), and ultimately controlling the turbulence morphology within the reactor pressure vessel to meet the target range of the core nuclear power fluctuation amplitude value, thereby eliminating or improving the impact of the core inlet turbulence on core power fluctuation, fuel assembly low-frequency oscillation, and grid wear.

[0094] It can be understood that the above embodiments only express some implementation methods of the present application, and their descriptions are relatively specific and detailed, but they cannot be understood as limiting the scope of the patent of the present application. It should be pointed out that for ordinary technicians in this field, without departing from the concept of the present application, the above technical features can be freely combined, and several deformations and improvements can be made, all of which fall within the scope of protection of the present application. Therefore, all equivalent changes and modifications made to the scope of the claims of the present application should fall within the scope of coverage of the claims of the present application.

Claims

1. A method for constructing and controlling the turbulent flow morphology of fluid inside a reactor pressure vessel, characterized in that: include: Step S1, establish a three-dimensional structural model of the reactor pressure vessel, and conduct an analysis or experiment on the flow field inside the reactor to obtain a dynamic flow field of the fluid inside the reactor, obtain relevant data of vortices and the time distribution characteristics and spatial distribution characteristics of turbulence in the inlet area of ​​each fuel assembly according to the dynamic flow field, and obtain the power spectrum density of turbulence in the inlet area of ​​each fuel assembly according to the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, perform energy spectrum analysis on the turbulence, and obtain the turbulence energy spectrum; Step S2, establish a three-dimensional flow field model of the fuel assembly, and according to the turbulence-related data, the time distribution characteristics and spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, and the turbulence energy spectrum, Through the low-frequency oscillation analysis of the fuel assembly, the change of the water gap of the fuel assembly is obtained, and according to the change of the water gap of the fuel assembly, it is determined whether the three-dimensional flow field model of the fuel assembly needs to be corrected, until the water gap of the fuel assembly does not change; step S3, establish a core calculation model, and according to the low-frequency oscillation analysis result of the fuel assembly, obtain the core nuclear power fluctuation amplitude value between each fuel assembly; step S4, determine whether the core nuclear power fluctuation amplitude value between each fuel assembly falls into the target interval, if not falling into the target interval, then change the internal three-dimensional structure of the reactor pressure vessel, repeat the steps S1 to S4, until the core nuclear power fluctuation amplitude value falls into the target interval.

2. The method for constructing and controlling the turbulent flow morphology of fluid inside a reactor pressure vessel according to claim 1, characterized in that: The step S1 comprises: step S11, establishing a three-dimensional structural model of the interior of the reactor pressure vessel, carrying out analysis or experiment of the flow field inside the reactor, and obtaining a dynamic flow field of the fluid inside the reactor; step S12, obtaining relevant data of the vortex by means of image recognition according to the dynamic flow field; step S13, extracting the average flow rate of each fuel assembly inlet area in a representative time period through the dynamic flow field, and obtaining the spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; step S14, extracting the change value of the flow rate in the inlet area of ​​each fuel assembly over time through the dynamic flow field, and obtaining the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; step S15, carrying out self-power spectrum analysis on the flow rate in the inlet area of ​​each fuel assembly according to the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly, and obtaining the power spectrum density of the turbulence in the inlet area of ​​each fuel assembly; step S16, carrying out energy spectrum analysis on the turbulence according to the scale of the vortex and the power spectrum density of the turbulence in the inlet area of ​​each fuel assembly, and obtaining the turbulence energy spectrum.

3. The method for constructing and controlling the turbulent flow pattern of the fluid inside the reactor pressure vessel according to claim 2, characterized in that: The step S13 comprises: extracting the average flow rate of each fuel assembly inlet area in a representative time period through the dynamic flow field, obtaining the flow distribution coefficient of each fuel assembly inlet area according to the average flow rate of each fuel assembly inlet area, and obtaining the maximum value, minimum value and overall spatial standard deviation value of the flow distribution of each fuel assembly inlet area according to the flow distribution coefficient, thereby obtaining the spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; And / or, the step S14 includes: extracting the time-varying value of the flow rate in the inlet area of ​​each fuel assembly through the dynamic flow field, obtaining the fluid pulsation value of the inlet area of ​​each fuel assembly based on the time-varying value of the flow rate in the inlet area of ​​each fuel assembly, and obtaining the standard deviation value of the flow rate in the inlet area of ​​each fuel assembly based on the fluid pulsation value, thereby obtaining the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly.

4. The method for constructing and controlling the turbulent flow morphology of fluid inside a reactor pressure vessel according to claim 1, characterized in that: In step S1, the vortex-related data includes the number, size, frequency and location of the vortex.

5. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 1, characterized in that: The step S2 includes: step S21, establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation and oscillation position of the fuel assembly through low-frequency oscillation analysis of the fuel assembly according to the relevant data of the vortex, the time distribution characteristics of turbulence in the inlet area of ​​each fuel assembly, and the spatial distribution characteristics; step S22, performing a fuel assembly water gap analysis according to the deformation and oscillation position of the fuel assembly to determine whether the fuel assembly water gap changes; if the fuel assembly water gap changes, adjusting the fuel assembly three-dimensional flow field model, and repeating the steps S21 and S22 until the fuel assembly water gap does not change.

6. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 5, characterized in that: The step S21 includes: step S211, establishing a three-dimensional flow field model with the core as the object, and obtaining the lateral flow distribution inside the core through fluid dynamics analysis based on the relevant data of the vortex, the time distribution characteristics of turbulence in the inlet area of ​​each fuel assembly, and the spatial distribution characteristics; step S212, obtaining the lateral loads on different positions of the fuel assembly based on the lateral flow distribution inside the core; step S213, obtaining the flow-induced vibration of the fuel assembly based on the lateral loads on different positions of the fuel assembly; step S214, obtaining the deformation and oscillation position of the fuel assembly based on the flow-induced vibration of the fuel assembly.

7. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 5, characterized in that: The step S21 includes: establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation and oscillation position of the fuel assembly through low-frequency oscillation analysis of the fuel assembly according to the relevant data of the vortex and the turbulence energy spectrum.

8. The method for constructing and controlling the turbulent flow morphology of fluid inside a reactor pressure vessel according to claim 1, characterized in that: The step S3 includes: step S31, establishing a core calculation model; step S32, obtaining the core neutron transport situation according to the change of the water gap of the fuel assembly; step S33, obtaining the heating power of the fuel assembly and the partition loading situation of the fuel assembly; step S34, obtaining the core nuclear power fluctuation amplitude value between each fuel assembly according to the heating power of the fuel assembly, the core neutron transport situation, the low-frequency oscillation analysis result of the fuel assembly and the partition loading situation of the fuel assembly.

9. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 8, characterized in that: The step S34 comprises: obtaining the core nuclear power fluctuation amplitude value between each fuel assembly through the core nuclear power fluctuation analysis of coupled nuclear-thermal-mechanical loads according to the heating power of the fuel assembly, the neutron transport condition of the core, the low-frequency oscillation analysis result of the fuel assembly and the partitioned loading condition of the fuel assembly.

10. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 8, characterized in that: The step S2 comprises: step S21, establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation and oscillation position of the fuel assembly through low-frequency oscillation analysis of the fuel assembly according to the relevant data of the vortex, the time distribution characteristics of turbulence in the inlet area of ​​each fuel assembly, and the spatial distribution characteristics; step S22, performing a fuel assembly water gap analysis according to the deformation and oscillation position of the fuel assembly to determine whether the fuel assembly water gap changes; if the fuel assembly water gap changes, adjusting the fuel assembly three-dimensional flow field model, and repeating the steps S21 and S22 until the fuel assembly water gap does not change; The step S34 includes: obtaining the core nuclear power fluctuation amplitude value between each fuel assembly according to the heating power of the fuel assembly, the core neutron transport condition, the oscillation position condition of the fuel assembly and the partition loading condition of the fuel assembly.

11. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 1, characterized in that: The step S4 comprises: step S41, judging whether the core nuclear power fluctuation amplitude value between each fuel assembly falls within the target range, if not, establishing the target limit range of the time distribution characteristics and the spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly based on the target range; step S42, judging whether the time distribution characteristics and the spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly obtained in the step S1 fall within the respective target limit ranges, if not, changing the internal three-dimensional structure of the reactor pressure vessel, repeating the step S1, until the time distribution characteristics and the spatial distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly obtained in the step S1 fall within the respective target limit ranges; step S43, repeating the steps S2 and S3, judging whether the core nuclear power fluctuation amplitude value measured in the step S3 this time falls within the target range; if not, re-changing the internal three-dimensional structure of the reactor pressure vessel, and repeating the steps S1, S42 and S43, until the core nuclear power fluctuation amplitude value is within the target range.

12. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 2, characterized in that: The step S1 further includes: step S17, obtaining a flow correlation coefficient between the inlet area of ​​each fuel assembly and the inlet areas of other fuel assemblies according to the dynamic flow field and the time distribution characteristics of the turbulence in the inlet area of ​​each fuel assembly; The step S2 comprises: step S21, establishing a three-dimensional flow field model with the core as the object, and obtaining the deformation, oscillation position, and low-frequency oscillation frequency of the fuel assembly through low-frequency oscillation analysis of the fuel assembly according to the relevant data of the vortex, the time distribution characteristics of turbulence in the inlet area of ​​each fuel assembly, and the spatial distribution characteristics; step S22, performing a fuel assembly water gap analysis according to the deformation and oscillation position of the fuel assembly to determine whether the fuel assembly water gap changes; if the fuel assembly water gap changes, adjusting the fuel assembly three-dimensional flow field model, and repeating the steps S21 and S22 until the fuel assembly water gap does not change; The step S4 comprises: step S411, judging whether the core nuclear power fluctuation amplitude values ​​between the fuel assemblies fall into the target interval; if not, performing multi-parameter fitting on the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, the turbulence energy spectrum, the low-frequency oscillation frequency, and the core nuclear power fluctuation amplitude value to obtain a multi-parameter fitting relationship; step S412, according to the multi-parameter fitting relationship, based on the target interval of the nuclear power fluctuation amplitude value, establishing a target limit range of the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum that meets the target interval condition; step S42, judging whether the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in step S1 The value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum fall into their respective target limit ranges; if not, the internal three-dimensional structure of the reactor pressure vessel is changed, and step S1 is repeated until the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in step S1 all fall into their respective target limit ranges; step S43, repeating steps S2 and S3, and judging whether the core nuclear power fluctuation amplitude value measured in step S3 this time falls into the target interval; if not, the internal three-dimensional structure of the reactor pressure vessel is changed again, and steps S1, S42, and S43 are repeated until the core nuclear power fluctuation amplitude value falls into the target interval.

13. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 12, characterized in that: The step S42 comprises: judging whether the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in the step S1 fall within their respective target limit ranges; if not, establishing at least one turbulence control region in the reactor pressure vessel, improving the dynamic flow field morphology in the reactor pressure vessel by adding a vortex breaking component in the at least one turbulence control region, and repeating the step S1 until the flow distribution coefficient, the fluid pulsation value, the power spectrum density, the flow correlation coefficient, and the turbulence energy spectrum measured in the step S1 all fall within their respective target limit ranges; The step S43 includes: repeating the step S2 and the step S3, and determining whether the core nuclear power fluctuation amplitude value measured in the step S3 falls within the target interval; if not, improving the dynamic flow field morphology in the reactor pressure vessel again, and repeating the step S1, the step S42 and the step S43 until the core nuclear power fluctuation amplitude value falls within the target interval.

14. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 13, characterized in that: The number of the at least one turbulence control area is three, namely, an annular descending section area, a lower chamber area and an upper chamber area.

15. The method for constructing and controlling the turbulent flow pattern of fluid inside a reactor pressure vessel according to claim 12, characterized in that: In step S17, the flow correlation coefficient is calculated by formula 1, and the calculation process of formula 1 is: Among them, C ab is the flow correlation coefficient between the inlet area of ​​fuel assembly a and the inlet area of ​​fuel assembly b; is the average value of the fluid pulsation values ​​of all fuel assembly inlet areas centered at the inlet area of ​​fuel assembly a; is the average value of the fluid pulsation values ​​of all fuel assembly inlet areas centered at the inlet area of ​​fuel assembly b; is the variance of the fluid pulsation values ​​at the inlet area of ​​all fuel assemblies centered at the inlet area of ​​fuel assembly a; is the variance of the fluid pulsation values ​​at the inlet area of ​​all fuel assemblies centered at the inlet area of ​​fuel assembly b.

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