Transient simulation method for nuclear reactor core

Abstract

This application provides a method for transient core simulation of a nuclear reactor. The method is applied to an electronic device equipped with a simulation platform. The simulation platform houses a thermal-hydraulic model and a core simulation model of the nuclear reactor. The thermal-hydraulic model sends target thermal parameters of the nuclear reactor at a first moment to the simulation platform. The simulation platform sends transient calculation commands and target thermal parameters to the core simulation model. Responding to the transient calculation commands, the core simulation model determines the transient core simulation result at the first moment based on the target thermal parameters and the neutron integral equilibrium equations of the nuclear reactor core. This application enables the core transient model to be instructed to perform core transient simulation via transient calculation commands, and the core transient model automatically performs transient calculations of the reactor core, thereby improving the efficiency of transient core simulation of nuclear reactors.

Description

Technical Field

[0001] This application relates to the field of artificial intelligence technology, and in particular to a method for transient core simulation of a nuclear reactor. Background Technology

[0002] In the design, operation and maintenance of nuclear reactors, core transient calculations are a key step in assessing the safety and performance of nuclear reactors. Core transient calculations involve the dynamic analysis of thermal-hydraulic parameters and neutron physics parameters of nuclear reactors under unsteady-state conditions.

[0003] However, existing nuclear reactor core transient calculations are generally performed by simulating the nuclear reactor using a simulation platform to obtain thermal-hydraulic parameters. Then, the core transient analysis software is manually operated to perform core transient simulation based on the obtained thermal-hydraulic parameters. This results in low efficiency of core transient analysis and the inability to obtain core transient simulation data in a timely manner. Summary of the Invention

[0004] The main objective of this application is to propose a method for transient simulation of nuclear reactor core, which aims to obtain transient simulation data of nuclear reactor core in a timely manner, thereby improving the efficiency of transient simulation of nuclear reactor core.

[0005] To achieve the above objectives, a first aspect of this application proposes a core transient simulation method for a nuclear reactor, applied to an electronic device. The electronic device is equipped with a simulation platform, which houses a thermal-hydraulic model and a core simulation model of the nuclear reactor. The method includes:

[0006] The target thermal parameters of the nuclear reactor at the first moment are sent to the simulation platform through the thermal-hydraulic model.

[0007] The simulation platform sends transient calculation commands for the nuclear reactor and the target thermal parameters to the core simulation model.

[0008] The core simulation model responds to the transient calculation command and, based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core, determines the core transient simulation result of the nuclear reactor at the first moment. The neutron integral equilibrium equation is spatially discretized using a nonlinear semi-analytical nodal method to determine the transient neutron diffusion equation, which is calculated using an implicit difference method. Furthermore, the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal.

[0009] To achieve the above objectives, a second aspect of this application provides a transient core simulation device for a nuclear reactor, applied to an electronic device. The electronic device is equipped with a simulation platform, on which a thermal-hydraulic model and a core simulation model of the nuclear reactor are built. The device includes:

[0010] The first transmitting module is used to transmit the target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal-hydraulic model.

[0011] The second sending module is used to send transient calculation commands for the nuclear reactor and the target thermal parameters to the core simulation model through the simulation platform;

[0012] The transient simulation module is used to respond to the transient calculation command through the core simulation model, and determine the core transient simulation result of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core; wherein, the neutron integral equilibrium equation is spatially discretized to determine the transient neutron diffusion equation through a nonlinear semi-analytical nodal method, the transient neutron diffusion equation is calculated through an implicit difference method, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal.

[0013] To achieve the above objectives, a third aspect of this application provides an electronic device, which includes a memory and a processor. The memory stores a computer program, and the processor executes the computer program to implement the method described in the first aspect.

[0014] To achieve the above objectives, a fourth aspect of the present application provides a computer-readable storage medium storing a computer program that, when executed by a processor, implements the method described in the first aspect.

[0015] This application proposes a core transient simulation method for nuclear reactors. This method is applied to electronic equipment equipped with a simulation platform. The simulation platform houses a thermo-hydraulic model and a core simulation model of the nuclear reactor. The thermo-hydraulic model sends the target thermo-hydraulic parameters of the nuclear reactor to the simulation platform at the first moment. The simulation platform sends transient calculation commands and the target thermo-hydraulic parameters to the core simulation model. The core simulation model responds to the transient calculation commands and, based on the target thermo-hydraulic parameters and the neutron integral equilibrium equations of the nuclear reactor core, determines the core transient simulation results at the first moment. Wherein, neutrons... The integral equilibrium equations are spatially discretized to determine the transient neutron diffusion equations using a nonlinear semi-analytical nodal approach. These transient neutron diffusion equations are calculated using an implicit difference method, and the average neutron source term within a nodal in the neutron integral equilibrium equations is determined by volume integration within the nodal. Thus, this invention provides a core transient model that can be instructed to perform core transient simulations via transient calculation commands. The core transient model responds to these commands, automatically and intelligently performing core transient calculations based on target thermal parameters and obtaining the results, thus avoiding manual operation and improving computational efficiency. Furthermore, obtaining the transient neutron diffusion equation through the implicit difference method can better handle the temporal variations of the main parameters in the reactor core, reducing numerical instability and time step limitations, thereby improving the accuracy of transient simulation of the reactor core. Then, spatial discretization of the transient neutron diffusion equation using the nonlinear semi-analytical nodal method can more accurately simulate neutron behavior within the nuclear reactor core. Therefore, obtaining the transient neutron diffusion equation through the implicit difference method and then spatially discretizing it using the nonlinear semi-analytical method yields a more accurate and efficient neutron integral equilibrium equation, further improving the accuracy of transient core simulation results and increasing the efficiency of transient reactor core simulation. Attached Figure Description

[0016] Figure 1 This is a flowchart of a nuclear reactor core transient simulation method provided in an embodiment of this application;

[0017] Figure 2 This is a schematic diagram of the structure of a transient simulation device for the core of a nuclear reactor provided in an embodiment of this application;

[0018] Figure 3 This is a schematic diagram of the hardware structure of the electronic device provided in the embodiments of this application. Detailed Implementation

[0019] To make the objectives, technical solutions, and advantages of this application clearer, the following detailed description is provided in conjunction with the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are merely illustrative and not intended to limit the scope of this application.

[0020] It should be noted that although functional modules are divided in the device schematic diagram and a logical order is shown in the flowchart, in some cases, the steps shown or described may be performed in a different order than the module division in the device or the order in the flowchart. The terms "first," "second," etc., in the specification, claims, and the aforementioned drawings are used to distinguish similar objects and are not necessarily used to describe a specific order or sequence.

[0021] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this application belongs. The terminology used herein is for the purpose of describing embodiments of this application only and is not intended to limit this application.

[0022] This application provides a method for transient simulation of nuclear reactor core, which aims to obtain transient simulation data of nuclear reactor core in a timely manner, thereby improving the efficiency of transient simulation of nuclear reactor core.

[0023] The nuclear reactor core transient simulation method provided in this application is specifically illustrated through the following embodiments. First, the nuclear reactor core transient simulation method in this application embodiment is described.

[0024] The nuclear reactor core transient simulation method provided in this application relates to the field of artificial intelligence technology. This method can be applied to a terminal, a server, or software running on either a terminal or a server. In some embodiments, the terminal can be an electronic device such as a smartphone, tablet, laptop, or desktop computer.

[0025] Figure 1 This is an optional flowchart of a nuclear reactor core transient simulation method provided in this application embodiment. The method is applied to an electronic device equipped with a simulation platform for simulating a nuclear reactor. In some embodiments, the simulation platform is equipped with a thermal-hydraulic model and a core simulation model of the nuclear reactor. The thermal-hydraulic model is used to obtain the thermal parameters of the nuclear reactor, which may include coolant temperature, effective fuel temperature, coolant density, or soluble poison density, etc. The core simulation model is used to perform transient simulation of the nuclear reactor.

[0026] Figure 1 The method may include, but is not limited to, steps S101 to S103, wherein:

[0027] Step S101: Send the target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal-hydraulic model.

[0028] In this step, at each preset calculation cycle (time step), the thermal-hydraulic model will automatically feed back thermal parameters to the simulation platform. The first moment is the latest moment (current moment), so as to enable the simulation platform to send transient calculation commands for the nuclear reactor and the target thermal parameters of the nuclear reactor at the latest moment to the core simulation model in real time.

[0029] Step S102: Send the target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal-hydraulic model.

[0030] In this step, when the core simulation model needs to run a transient simulation of the nuclear reactor at the first moment according to the requirements, the operator can select an operation command on the operation interface to control the simulation platform. This operation command is used to instruct the simulation platform to send transient calculation commands for the nuclear reactor and the target thermal parameters of the nuclear reactor at the first moment to the core simulation model. In response to the operation command, the simulation platform sends transient calculation commands for the nuclear reactor and the target thermal parameters of the nuclear reactor at the first moment to the core simulation model.

[0031] In some implementations, the simulation platform sends the calculation results of the previous time step to the core simulation model. The calculation results include, but are not limited to, the neutron flux density and the concentration of fission products at the previous time step. The fission products include fission product 135I, fission product 135Xe, fission product 149Pm and fission product 149Sm.

[0032] Step S103: Responding to the transient calculation command through the core simulation model, the transient simulation result of the nuclear reactor core at the first moment is determined based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core; wherein, the neutron integral equilibrium equation is spatially discretized to determine the transient neutron diffusion equation through a nonlinear semi-analytical nodal method, the transient neutron diffusion equation is calculated through an implicit difference method, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal.

[0033] In this step, a core transient model is pre-built. The core transient model receives a reset command from the simulation platform and performs a reset (initial load) based on the reset command, establishes a communication interface with the simulation platform, and reads the model input card. The core transient model is initialized based on the model input card to obtain the core simulation model of the nuclear reactor. The transient calculation command is used to instruct the core simulation model to perform transient simulation of the nuclear reactor core. The model input card includes, but is not limited to, core layout, node size, and node division.

[0034] The aforementioned communication interface is used to receive control commands and thermal parameters sent in real time by the simulation platform. The control commands instruct the core transient model to perform corresponding operations. These control commands include freeze commands, run commands (transient calculation commands), store commands, or backtrack commands. The freeze command instructs the core transient model to perform no processing; the store command instructs the core transient model to output its calculation results to an IC file; and the backtrack command instructs the core transient model to execute a run backtrack function. For example, the simulation platform sends transient calculation commands and the target thermal parameters of the nuclear reactor at the first moment to the core simulation model through the communication interface connected to the core simulation model.

[0035] The above-mentioned neutron integral equilibrium equations spatially discretize the transient neutron diffusion equations using a nonlinear semi-analytical nodal method. Specifically, the transient neutron diffusion equations temporally discretize the spacetime neutron dynamics equations using an implicit difference method, and the nodal average neutron source term in the neutron integral equilibrium equations is determined by volume integration within the nodal.

[0036] The core simulation model receives transient calculation commands from the simulation platform via a communication interface. In response to the transient calculation commands, it calculates the core transient simulation results of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core.

[0037] Steps S101 to S103, as illustrated in this embodiment, send transient calculation commands for the nuclear reactor to the core simulation model via the simulation platform. This allows the core simulation model to intelligently execute the core transient simulation function in response to the transient calculation commands, eliminating the need for manual core transient simulation operations and improving the efficiency of core transient simulation. Secondly, obtaining the transient neutron diffusion equation through the implicit difference method better handles the temporal variations of key parameters in the reactor core, reducing numerical instability and time step limitations, thereby improving the accuracy of the reactor core transient simulation. Furthermore, spatial discretization of the transient neutron diffusion equation using a nonlinear semi-analytical nodal method further refines the simulation. To accurately simulate neutron behavior within the reactor core, the transient neutron diffusion equation is obtained through an implicit finite difference method. This equation is then spatially discretized using a nonlinear semi-analytical method, resulting in a more accurate and efficient neutron integral equilibrium equation. This further improves the accuracy and efficiency of transient core simulation results. Furthermore, the simulation platform can send transient calculation commands in real-time, instructing the core simulation model to perform transient core simulations based on the latest thermal parameters. The calculation speed meets real-time requirements, and the process of this transient core simulation method can accommodate various operating conditions, making it more widely applicable.

[0038] In some embodiments, a transient neutron diffusion equation is constructed before step S103, and the steps before step S103 may include, but are not limited to, steps S201 to S207.

[0039] Step S201: Use the time derivative of the neutron flux density as the first parameter.

[0040] Step S202: The ratio of the first parameter to the neutron flux density is used as the first dynamic frequency.

[0041] Step S203: Use the time derivative of the delayed neutron precursor nuclear density as the second parameter.

[0042] Step S204: The ratio of the second parameter to the neutron precursor nuclear density is used as the second dynamic frequency.

[0043] Step S205: Discretize the first dynamic frequency and the second dynamic frequency respectively using implicit differential processing to obtain the differential representation of the first dynamic frequency and the differential representation of the second dynamic frequency.

[0044] Step S206: Based on the differential representation of the first dynamic frequency and the differential representation of the second dynamic frequency, the transient neutron diffusion equation is obtained.

[0045] Step S207: Spatial discretize the transient neutron diffusion equation to obtain the neutron integral equilibrium equation.

[0046] In this implementation, the reactor core is divided into multiple segments according to the Cartesian coordinate system. Dynamic frequency expressions for the neutron flux density and the delayed neutron precursor nucleus density of each segment as a function of time are defined. The time terms in the dynamic frequency expressions for neutron flux density (i.e., the first dynamic frequency representation) and delayed neutron precursor nucleus density (i.e., the second dynamic frequency representation) are discretized using the implicit difference method to obtain the difference representations of the dynamic frequency of neutron flux density and delayed neutron precursor nucleus density. These difference representations are then substituted into the spacetime neutron dynamics equations to eliminate the delayed precursor neutron nucleus concentration, resulting in the transient neutron diffusion equation. Finally, the transient neutron diffusion equation is spatially discretized to obtain the neutron integral balance equation.

[0047] Specifically, the dynamic frequency expressions for neutron flux density and delayed neutron precursor nuclear density can be represented by the following equation (1):

[0048]

[0049] Where, ω g,ijk φ represents the dynamic frequency of the neutron flux density of the g-group in the ijk-node. g,ijk Let represent the neutron flux density of the g-group in the ijk section, and t represent the running time of the transient model. Let represent the dynamic frequency of the density of the m-th delayed neutron precursor nuclei in the ijk-th block, where ijk represents the block, g represents the neutron energy group, and m represents the number of delayed neutron precursor nuclei. This represents the density of the delayed neutron precursor nuclei in the m-th group.

[0050] In this embodiment, Let it be the first parameter, and Let it be denoted as the first dynamic frequency, and Let it be the second parameter. This is denoted as the second dynamic frequency.

[0051] By discretizing the time terms in the dynamic frequency expressions of neutron flux density and delayed neutron precursor nuclear density using an implicit difference method, the difference representations of the dynamic frequency of neutron flux density and delayed neutron precursor nuclear density can be obtained. These difference representations can be expressed by the following equation (2):

[0052]

[0053] Among them, t n+1 Indicates the current running time, t n It indicates the moment before the current running time.

[0054] Substituting the difference expression of the dynamic frequency of neutron flux density and the difference expression of the dynamic frequency of delayed neutron precursor nucleus density into the spacetime neutron dynamics equations, and eliminating the delayed precursor neutron nucleus concentration, we obtain the transient neutron diffusion equation, which can be expressed as follows (3):

[0055]

[0056] Among them, D g Indicates the diffusion coefficient. Q represents the macroscopic neutron removal cross section. g,ijk This represents the neutron source term.

[0057] Furthermore, the calculation formula for the above neutron source term can be expressed as follows (4):

[0058]

[0059] Where G represents the group number of the neutron energy group, χ represents the macroscopic scattering cross section of neutrons from the g′ energy group to the g energy group. g,ijk This represents the proportion of transient neutrons generated by fission into the g neutron energy group in the ijk node. φ represents the macroscopic fission cross section of the g′ neutron energy group in the ijk block. g′,ijk Let g' represent the neutron flux density of the g' neutron energy group in the ijk nodal, where g' represents the g' neutron energy group.

[0060] The calculation expressions for the macroscopic neutron removal cross section of the g neutron energy group in the ijk node and the instantaneous neutron fraction generated by fission into the g neutron energy group in the ijk node are redefined. The above expressions can be expressed by the following equation (5):

[0061]

[0062] in, This represents the macroscopic neutron fission cross section of the g neutron energy group in the redefined ijk section. V represents the macroscopic neutron fission cross section of the g neutron energy group in the ijk block. g The value represents the neutron velocity, and β represents the delayed neutron fraction of each group. m The sum, β m λ represents the proportion of the m-th delayed neutron group in all fission neutrons. m Denotes the decay constant of the m-th slow-emitting neutron precursor nucleus. This represents the proportion of slow-emitting neutrons produced by the m-th precursor nucleus into the g-group. This represents the instantaneous neutron fraction generated by the redefined fission into the g neutron energy group in the ijk block.

[0063] By redefining the macroscopic neutron removal cross section of the g neutron energy group in the ijk block and the calculation expression of the transient neutron fraction generated by fission to the g neutron energy group in the ijk block, the transient neutron diffusion equation of equation (3) is rearranged into the same expression as the steady-state neutron diffusion equation, so that the transient neutron diffusion equation can be spatially discretized using the semi-analytical nodal method to obtain the neutron integral equilibrium equation.

[0064] It should be noted that neutron energy groups can be divided into hot groups and fast groups, so G can be 2. Since the density of each group of delayed neutron precursor nuclei is calculated based on the macroscopic fission cross section and average neutron flux density of the neutron energy group of the same block, in this embodiment, the density of each group of delayed neutron precursor nuclei is recorded as the delayed neutron precursor nuclei of the corresponding block.

[0065] It should be noted that each node of the reactor core includes multiple neutron energy groups and multiple sets of delayed neutron precursor nuclei.

[0066] In this embodiment, by defining dynamic frequency expressions for neutron flux density and neutron precursor nuclei, and then using an implicit difference method to differentiate the time term in the above dynamic frequency expressions, difference expressions for neutron flux density and neutron precursor nuclei are obtained. Substituting the above difference expressions into the spacetime neutron dynamics equations yields the transient neutron diffusion equations, which can better handle the time-varying changes of the main parameters in the reactor core, reduce numerical instability and time step limitations, and improve the transient simulation accuracy of the reactor core.

[0067] In some embodiments, the above-mentioned neutron integral balance equation is used to characterize the relationship between the first average neutron flux density of the neutron energy group of the first block at the first time, the second average neutron flux density of the neutron energy group of the second block at the first time, and the average neutron source term within the neutron energy group of the first block; wherein, the second block is the adjacent block of the first block in each coordinate direction. In some embodiments, the neutron integral balance equation is established before step S103, and the steps before step S103 may include, but are not limited to, steps S301 to S303.

[0068] Step S301: For each second segment, determine the first coefficient corresponding to the second average neutron flux density of the second segment based on the target parameters of the first segment and the second segment in the target coordinate direction corresponding to the second segment, and the coupling correction factor of the first segment in the target coordinate direction; wherein, the target parameters in the target coordinate direction include the first size and diffusion coefficient of the first segment in the target coordinate direction, and the second size of the second segment in the target coordinate direction.

[0069] Step S302: Determine the second coefficient corresponding to the first average neutron flux density of the first block based on the first coefficient corresponding to the second average neutron flux density of each second block and the macroscopic neutron removal cross section of the neutron energy group.

[0070] Step S303 determines the neutron integral balance equation based on each of the first coefficient, the second coefficient, the first average neutron flux density at the first time, the second average neutron flux density at the first time, and the average neutron source term within the block.

[0071] In this implementation, the aforementioned neutron integral balance equation is used to calculate the first average neutron flux density of the neutron energy group of the first block at the first time step.

[0072] The target parameters mentioned above include the first size and diffusion coefficient of the first block in the target coordinate direction. The target coordinate direction is the coordinate direction corresponding to the second block. For example, if the second block is the block preceding the first block on the x-axis, then the target coordinate direction is the positive direction of the x-axis.

[0073] In step S201 of some embodiments, a first coefficient corresponding to the second average neutron flux density of each second block is determined. Specifically, for the second block corresponding to the target coordinate direction in the positive direction, the expression for the first coefficient is as follows:

[0074]

[0075] In the formula, Cf1 represents the first coefficient, h represents the node size, and D represents the diffusion coefficient. The coupling correction factor is represented by g, the neutron energy group is represented by u, the target coordinate axis corresponding to the target coordinate direction (u can be x, y, or z), u+ represents the target coordinate direction (i.e., the positive direction of the u-axis), and h is the target coordinate direction. u This represents the first dimension of the first block on the target coordinate axis corresponding to the target coordinate direction u+. h' represents the diffusion coefficient of the neutron energy group along the target coordinate direction u+. u This represents the second dimension of the second block on the target coordinate axis corresponding to the target coordinate direction u+. This represents the coupling correction factor of the neutron energy group in the target coordinate direction u+.

[0076] For the first coefficient of the second block corresponding to the negative target coordinate direction, the expression for the first coefficient is as follows:

[0077]

[0078] In the formula, Cf1 represents the first coefficient, h represents the node size, and D represents the diffusion coefficient. The coupling correction factor is represented by g, the neutron energy group is represented by u, the target coordinate axis corresponding to the target coordinate direction (u can be x, y, or z), u- represents the target coordinate direction (i.e., the negative direction of the target coordinate axis), and h is represented by g. u This represents the first dimension of the first block on the target coordinate axis corresponding to the target coordinate direction u. h' represents the diffusion coefficient of the g neutron energy group along the target coordinate direction u-. u This represents the second dimension of the second block on the target coordinate axis corresponding to the target coordinate direction u. This represents the coupling correction factor of the g neutron energy group in the target coordinate direction u-.

[0079] In step S202 of some embodiments, the second coefficient corresponding to the first average neutron flux density of the first block can be determined based on the first coefficient corresponding to the second average neutron flux density of each second block and the sum of the macroscopic neutron removal cross sections of the neutron energy group.

[0080] In one implementation, the expression for the second coefficient can be as follows:

[0081]

[0082] Where Cf2 represents the second coefficient, ijk indicates the first block, i+1jk indicates the block preceding the first block on the x-axis, i-1jk indicates the block following the first block on the x-axis, ij+1k indicates the block preceding the first block on the y-axis, ij-1k indicates the block following the first block on the y-axis, ijk+1 indicates the block preceding the first block on the z-axis, and ijk-1 indicates the block following the first block on the z-axis. This represents the macroscopic neutron removal section of the neutron energy group in the first block.

[0083] In step S203 of some embodiments, a neutron integral equilibrium equation is constructed such that the sum of the first product and each of the second products equals the intra-block average neutron source term of the neutron energy group. Here, the first product is the product of the first coefficient and the first average neutron flux density at the first time step, and the second product is the product of the second coefficient and the corresponding second average neutron flux density at the first time step.

[0084] In one embodiment, the above-mentioned neutron integral equilibrium equation can be expressed by the following equation (6):

[0085]

[0086] in, This represents the first average neutron flux density. This represents the second average neutron flux density of the neutron energy group of the block preceding the first block on the x-axis. This represents the second average neutron flux density of the neutron energy group of the next block in the first block on the x-coordinate axis. This represents the second average neutron flux density of the neutron energy group of the block preceding the first block on the y-coordinate axis. This represents the second average neutron flux density of the neutron energy group of the next block in the first block on the y-coordinate axis. This represents the second average neutron flux density of the neutron energy group of the block preceding the first block on the z-coordinate axis. This represents the second average neutron flux density of the neutron energy group of the next block in the first block along the z-coordinate axis. t represents the intra-block average neutron source term of the neutron energy group in the first block. n+1 This indicates the current running time, i.e., the first moment.

[0087] The formula for calculating the average neutron source term within the above-mentioned block can be expressed by the following equation (7):

[0088]

[0089] Among them, Q g,ijk This represents the neutron source term.

[0090] In some embodiments, steps S401 to S404 may be included before step S301.

[0091] Step S401: Based on the diffusion coefficient of the first segment in the target coordinate direction, and the characteristic parameters of the first segment and the second segment corresponding to the target coordinate direction, determine the net neutron differential flux of the neutron energy group on the surface of the first segment in the target coordinate direction; wherein, the characteristic parameters include the average neutron flux density and size.

[0092] Step S402: Determine the net neutron correction flow of the neutron energy group on the surface of the first segment in the target coordinate direction based on the net neutron differential flow and the net neutron flow of the neutron energy group on the surface of the first segment in the target coordinate direction.

[0093] Step S403: Based on the net neutron correction flow and the characteristic parameters of the first block and the second block corresponding to the target coordinate direction, determine the coupling correction factor of the first block in the target coordinate direction.

[0094] The aforementioned characteristic parameters include average neutron flux density and size.

[0095] In this embodiment, the transverse integration method and the residual weight method can be used to solve for the net neutron flow and its correction flow, and then solve for the new coupling correction factor.

[0096] In step S401 of some embodiments, the expression for the net neutron differential current of the neutron energy group on the block surface of the first block in the target coordinate direction can be shown in the following equation (8):

[0097]

[0098] in, This represents the net neutron differential split of the neutron energy group on the surface of the first segment in the target coordinate direction. This represents the average neutron flux density of the neutron energy group in the second block corresponding to the target coordinate direction.

[0099] In step S402 of some embodiments, the expression for the net neutron correction flow of the neutron energy group on the block surface of the first block in the target coordinate direction is as follows: (9)

[0100]

[0101] in, This represents the net neutron correction flux of the neutron energy group on the surface of the first segment in the target coordinate direction. This represents the net neutron flux of the neutron energy group on the surface of the first segment along the target coordinate direction. This represents the net neutron differential current of the neutron energy group on the surface of the first segment in the target coordinate direction.

[0102] In step S403 of some embodiments, the expression for the coupling correction factor of the first block in the target coordinate direction is as shown in equation (10):

[0103]

[0104] in, This represents the coupling correction factor of the first block in the target coordinate direction.

[0105] In some embodiments, the core transient simulation results include the first average neutron flux density of the first neutron energy group of the first target block at a first time. The first target block is any block in the core of the nuclear reactor. The first target block includes multiple neutron energy groups. The first neutron energy group is any neutron energy group of the first target block. Step S103 may include, but is not limited to, steps S501 to S503.

[0106] Step S501: Based on the target thermal parameters at the first moment, determine the first cross-sectional data of the macroscopic neutron removal cross section of the first neutron energy group at the first moment.

[0107] Step S502: Obtain the coupling correction factor of the first neutron energy group in each coordinate direction at the first time point, the target parameter, and the average neutron source term within the block.

[0108] Step S503: Input the first cross-sectional data of the first neutron energy group at the first time, the coupling correction factor of the first neutron energy group in each coordinate direction at the first time, the target parameter, and the average neutron source term within the block into the neutron integral balance equation to obtain the first average neutron flux density at the first time.

[0109] In step S401 of some embodiments, this embodiment provides a calculation formula for calculating cross-sectional data of various cross-sectional types. The cross-sectional types can be divided into neutron macroscopic removal cross-sections, macroscopic fission cross-sections, etc. First, the cross-sectional data before control rod correction is calculated, and then the cross-sectional data before control rod correction is updated according to the position of the control rod to obtain the final cross-sectional data of the neutron energy group.

[0110] The formula for calculating the cross-sectional data of the control rod before correction is shown in equation (11):

[0111]

[0112] In the formula, ijk indicates the nodal; n represents the power of the expansion term; X indicates the cross-section type; g indicates the neutron energy group; This indicates the cross-sectional data of the control rod before correction; The Taylor expansion factor of the X-type section; ΔT c , Δρ C ΔN sp These represent the changes in coolant temperature, effective fuel temperature, coolant density, and soluble toxic substance (boric acid) density relative to the reference operating condition.

[0113] Then, based on the rod position update of each group of control rods and the cross-sectional data calculated using the above formula (11) at the first moment, the final cross-sectional data of the neutron energy group is obtained, where the cross-sectional data is calculated by the rod position update formula (11) of each group of control rods using the following formula (12):

[0114]

[0115] In the formula, This indicates the corrected cross-sectional data for the control bar; This represents the cross-section of the control rod before correction. This indicates the cross-sectional change introduced when the control rod is fully inserted into the segment; This represents the weighting factor of the control bar insertion block.

[0116] It should be noted that if the cross-section type is the macroscopic absorption cross-section of the fission products of the neutron energy group, the corrected macroscopic absorption cross-section of the fission products of the neutron energy group is calculated according to equations (11) and (12), and then the cross-sectional data of the macroscopic absorption cross-section at the first moment is updated according to the calculation results of the fission products (135Xe, 149Sm) at the previous moment by the following equation (13):

[0117]

[0118] In the formula, This represents the macroscopic absorption cross section after modification of the fission products. This represents the macroscopic absorption cross section before the fission products are corrected. The microscopic absorption cross sections of 135Xe and 149Sm at the previous moment are respectively; N Xe N Sm The densities are 135Xe and 149Sm, respectively.

[0119] In this embodiment, the cross-sectional data of different cross-sectional types of each neutron energy group in each block of the nuclear reactor core can be calculated by the above calculation formulas (11) and (12).

[0120] In this embodiment, any segment in the core of the nuclear reactor is designated as the first target segment, and any neutron energy group of the first target segment is designated as the first neutron energy group. Steps S501-S502 are explained with the first target segment and the first neutron energy group as the objects. Specifically, according to the above-mentioned cross-sectional data calculation formulas (11) and (12), the first cross-sectional data of the neutron macroscopic removal cross-section of the first neutron energy group of the first target segment is determined based on the target thermal parameters.

[0121] In step S502 of some embodiments, the coupling correction factor of the first neutron energy group in each coordinate direction can be obtained by equations (8), (9), and (10) in steps S401-S403. Since the first average neutron flux density of the neutron energy group is used as a parameter in the expression of the coupling correction factor, the coupling correction factor of the first neutron energy group in each coordinate direction at the current time (i.e., the first time) can be calculated based on the block average neutron flux density of the first neutron energy group in each coordinate direction at the previous time (i.e., the second time).

[0122] In step S503 of some embodiments, the first cross-sectional data of the first neutron energy group at the first time moment, the coupling correction factor of the first neutron energy group at the first time moment in each coordinate direction, the target parameters in each coordinate direction, and the average neutron source term within the block are input into the neutron integral balance equation. A neutron integral balance equation with only the average neutron flux density at the current time as the unknown can be obtained. The neutron integral balance equation is solved to obtain the first average neutron flux density of the first neutron energy group at the current time.

[0123] In some embodiments, the core transient simulation results include the slow-emission neutron precursor nuclear density of the first target block at the first moment, and step S103 may include, but is not limited to, steps S601 to S603.

[0124] Step S601: For each neutron energy group of the first target block, based on the target thermal parameters at the second time and the neutron integral balance equation of the reactor core, determine the first average neutron flux density of the neutron energy group of the first target block at the second time, where the second time is the time before the first time.

[0125] Step S602: For each neutron energy group of the first target block, determine the second cross-sectional data of the macroscopic fission cross-section of the neutron energy group at the second time according to the target thermal parameters at the second time; and determine the first target value of the neutron energy group according to the second cross-sectional data of the neutron energy group and the first average neutron flux density of the neutron energy group of the first target block at the second time.

[0126] Step S603: Determine the slowed neutron precursor nuclear density of the first target block at the first time step based on the first target value of each of the neutron energy groups and the slowed neutron precursor nuclear density of the first target block at the second time step.

[0127] This embodiment provides a method for calculating the nuclear density of each group of delayed neutron precursors, and the calculation formula can be expressed by the following equation (14):

[0128]

[0129] In the formula, C m Let represent the nuclear density of the delayed neutron precursor in the m-th group, and G be the group number of the neutron energy group. This represents the macroscopic fission cross section of the g′ neutron energy group. β represents the average neutron flux density of the g′ neutron energy group in the ijk node. m λ represents the proportion of the m-th delayed neutron group in all fission neutrons. m Let a represent the decay constant of the m-th delayed neutron precursor nucleus. 1m and a 2m Represents the coefficient.

[0130] Among them, a 1m and a 2m It can be calculated using the following formula (15):

[0131]

[0132] Where, Δt n It represents the size of a time step (i.e., the time difference between the second time step and the first time step).

[0133] In this embodiment, the neutron energy group can be divided into hot and fast groups, so G can be 2. Since the density of each group of delayed neutron precursor nuclei is calculated based on the macroscopic fission cross section and average neutron flux density of the neutron energy group of the same block, in this embodiment, the density of each group of delayed neutron precursor nuclei is recorded as the density of the delayed neutron precursor nuclei of the corresponding block.

[0134] In this embodiment, This is denoted as the first target value.

[0135] In step S601 of some embodiments, firstly, the first average neutron flux density of each neutron energy group of the first target block at the previous time (i.e., the second time) at the current time is calculated according to the above steps S501-S503.

[0136] In step S602 of some embodiments, the second cross-sectional data of the macroscopic fission cross-section of each neutron energy group of the first target block at the second time is calculated according to the calculation formulas (11) and (12) of the cross-sectional data. For each neutron energy group, the product between the corresponding second cross-sectional data at the second time and the first average neutron flux density at the second time is calculated to obtain the first target value of the neutron energy group.

[0137] In step S603 of some embodiments, the proportion of the m-th group of delayed neutrons corresponding to the first target block in all fission neutrons and the decay constant of the m-th group of delayed neutron precursor nuclei are obtained. According to the above formula (14), based on the proportion of the m-th group of delayed neutrons corresponding to the first target block in all fission neutrons and the decay constant of the m-th group of delayed neutron precursor nuclei, the first target value of each neutron energy group corresponding to the first target block, the first average neutron flux density of the neutron energy group of the first target block at the second time, and the first coefficient a 1m Second coefficient a 2m Calculate the slowed neutron precursor nuclear density of the first target block.

[0138] In some embodiments, the transient simulation results of the reactor core include the concentration of fission products of the first target node at the first time point, wherein the first target node is any node in the reactor core of the nuclear reactor, the first target node includes multiple neutron energy groups, and the fission products include at least one of fission product 135I, fission product 135Xe, fission product 149Pm, and fission product 149Sm. Step S103 includes, but is not limited to, steps S701 to S705.

[0139] Step S701: Obtain the concentration of fission products of the first target block at the second time point, where the second time point is the time point preceding the first time point;

[0140] Step S702: For each neutron energy group of the first target block, based on the target thermal parameters at the first moment and the neutron integral balance equation of the nuclear reactor, determine the first average neutron flux density of the neutron energy group of the first target block at the first moment.

[0141] Step S703: For each neutron energy group of the first target block, determine the cross-sectional data of the macroscopic absorption cross section of the fission products of the neutron energy group according to the target thermal parameters at the first moment, and determine the second target value of the fission products of the neutron energy group at the first moment according to the cross-sectional data of the macroscopic absorption cross section of the fission products of the neutron energy group and the first average neutron flux density of the neutron energy group at the first moment.

[0142] Step S704: Obtain the concentration change of the fission products from the second time point to the first time point, wherein the concentration change is determined by target information, the target information including relevant parameters of the fission products and the concentration of the fission products in the first target block at the second time point, the relevant parameters of the fission products including the fission yield and decay constant of the fission products; if the fission products are 135I or 149Pm, then the target information also includes a second target value of the fission products of each neutron energy group in the first target block at the first time point; if the fission products are 135Xe, then the target information also includes a third target value, a fourth target value, and the second target value of the fission products of each neutron energy group in the first target block at the first time point, the third target value... The fourth target value is determined based on the decay constant of 135I and the concentration of 135I at the first time. The fourth target value is determined based on the cross-sectional data of the microscopic absorption cross-section of 135Xe of each neutron energy group of the first target block at the first time, and the first average neutron flux density of each neutron energy group at the first time. If the fission product is 149Sm, the target information also includes a fifth target value and a sixth target value. The fifth target value is determined based on the decay constant of 149Pm and the concentration of 149Pm at the first time. The sixth target value is determined based on the cross-sectional data of the microscopic absorption cross-section of 149Sm of each neutron energy group of the first target block at the first time, and the first average neutron flux density of each neutron energy group at the first time.

[0143] Step S705: Determine the concentration of fission products of the first target block at the first time based on the concentration of fission products of the first target block at the second time and the change in concentration of the fission products from the second time to the first time.

[0144] In this embodiment, a formula for calculating the concentration of fission products in each segment is provided, wherein the formula for calculating the concentration of fission product 135I at the first moment is as follows (16):

[0145]

[0146] in, Indicates the concentration of fission product 135I in the nodule, γ I λ represents the fission yield of fission product 135I. I The decay constant of the fission product 135I is represented. This represents the macroscopic absorption cross section of the fission product 135I of the g′ group in the ijk nodal. t represents the average neutron flux density of the g′ group in the ijk node. n t represents the time before the current time (i.e., the second time). n+1 Indicates the current time, Δtn It represents a time step (i.e., from the second time point to the first time point).

[0147] The formula for calculating the concentration of the fission product 135Xe at the first moment is shown in equation (17):

[0148]

[0149] in, This indicates the concentration of 135Xe, a fission product, in the ijk node. I Indicates the fission yield of fission product 135I, γ Xe λ represents the fission yield of the fission product 135Xe. Xe This represents the decay constant of the fission product 135Xe. This represents the macroscopic absorption cross section of the fission product 135Xe of the g′ neutron energy group in the ijk block. This represents the average neutron flux density of the neutron energy group in the ijk node g′. The value represents the microscopic absorption cross section of the fission product 135Xe of the g′ neutron energy group in the ijk block at the first moment.

[0150] The formula for calculating the concentration of fission products 149Pm at the first moment is shown in equation (18):

[0151]

[0152] in, This indicates the concentration of fission products 149Pm in the ijk node, γ Pm λ represents the fission yield of the fission product 149Pm. Pm This represents the decay constant of the fission product 149Pm. This represents the macroscopic absorption cross section of the fission product 149Pm of the g′ neutron energy group in the ijk block. Let represent the average neutron flux density of the g′ neutron energy group in the ijk section.

[0153] The formula for calculating the concentration of fission product 149Sm at the first moment is shown in equation (19):

[0154]

[0155] in, This indicates the concentration of 149Sm, a fission product, in the ijk node. Pm This represents the fission yield of 149 Pm of fission products. This represents the macroscopic absorption cross section of the fission product 149Sm of the g′ neutron energy group in the ijk block. This represents the average neutron flux density of the neutron energy group in the ijk node g′. The value represents the microscopic absorption cross section of the fission product 149Sm of the g′ neutron energy group in the ijk block at the first moment.

[0156] In this embodiment, Let this be the second target value. This is denoted as the third objective value. Let this be the fourth objective value. This is denoted as the fifth target value. This is denoted as the sixth target value.

[0157] Therefore, for the concentration calculation formula of fission product 135I or fission product 149Pm, the change in the concentration of fission product from the second time to the first time is determined based on the relevant parameters of the fission product, the concentration of the fission product of the first target block at the second time, and the second target value of the fission product of each neutron energy group of the first target block at the first time. The relevant parameters of the fission product include the fission yield and decay constant of the fission product.

[0158] The formula for calculating the concentration of the fission product 135Xe is based on the following: the change in the concentration of the fission product from the second time step to the first time step is determined according to the relevant parameters of 135Xe, the concentration of 135Xe in the first target block at the second time step, the decay constant of 135I, the concentration of 135I at the first time step, the cross-sectional data of the microscopic absorption cross-section of 135Xe in each neutron energy group of the first target block at the first time step, the first average neutron flux density of each neutron energy group at the first time step, and the second target value at the first time step.

[0159] The formula for calculating the concentration of fission product 149Sm is based on the relevant parameters of 149Sm, the concentration of 149Sm in the first target block at the second time, the decay constant of 149Pm, the concentration of 149Pm at the first time, the cross-sectional data of the microscopic absorption cross-section of 149Sm in each neutron energy group of the first target block at the first time, and the first average neutron flux density of each neutron energy group at the first time.

[0160] In step S701 of some embodiments, the concentration of fission products in the previous time step is obtained.

[0161] In step S702 of some embodiments, the first average neutron flux density of each neutron energy group of the first target block at the first time moment is calculated according to the above steps S501-S503.

[0162] In step S703 of some embodiments, the cross-sectional data of the macroscopic absorption cross-section of the fission products of each neutron energy group of the first target block are calculated according to the calculation formulas (11), (12), and (13) of the cross-sectional data described above. For each neutron energy group, the cross-sectional data of the macroscopic absorption cross-section corresponding to the fission products of each neutron energy group is calculated as a product of the first average neutron flux density to obtain the second target value of the fission products of the neutron energy group.

[0163] In step S704 of some embodiments, the change in concentration of fission products from the second time point to the first time point is calculated.

[0164] In step S705 of some embodiments, the concentration of fission products of the first target block at the first time is calculated according to the fission product concentration calculation formula, based on the concentration of fission products of the first target block at the second time and the change in the concentration of fission products from the second time to the first time.

[0165] In some embodiments, the core transient simulation results include the fission power of a first target block, which is any block in the core of the nuclear reactor. The first target block includes multiple neutron energy groups. Step S103 may include, but is not limited to, steps S801 to S803.

[0166] Step S801: For each neutron energy group of the first target block, based on the target thermal parameters at the first moment and the neutron integral balance equation of the reactor core, determine the first average neutron flux density of the neutron energy group of the first target block at the first moment.

[0167] Step S802: For each neutron energy group of the first target block, determine the cross-sectional data of the macroscopic energy generation section of the neutron energy group according to the target thermal parameters at the first moment; and determine the seventh target value of the neutron energy group according to the cross-sectional data of the macroscopic energy generation section of the neutron energy group, the first average neutron flux density of the neutron energy group of the first target block at the first moment, and the volume of the first target block.

[0168] Step S803: Determine the fission power of the first target block at the first moment based on the seventh target value of each of the neutron energy groups of the first target block.

[0169] In this embodiment, a method for calculating the fission power of a segment is provided, and the calculation formula is shown in the following equation (20):

[0170]

[0171] in, This represents the fission power of the g′ neutron energy group in the ijk section. This represents the macroscopic energy production cross section of the g′ neutron energy group in the ijk section. V represents the average neutron flux density of the g′ neutron energy group in the ijk node. ijk t represents the volume of the ijk nodules. n+1 This indicates the current moment, or the first moment.

[0172] In this embodiment, This is denoted as the seventh target value.

[0173] In step S801 of some embodiments, the first average neutron flux density of each neutron energy group of the first target block at the first time moment is calculated according to steps S501-S503 above.

[0174] In step S802 of some embodiments, the cross-sectional data of the macroscopic energy generation cross-section of each neutron energy group of the first target block at the first moment is calculated according to the calculation formulas (11) and (12) of the cross-sectional data described above. For each neutron energy group, the cross-sectional data of the macroscopic energy generation cross-section corresponding to the neutron energy group, the first average neutron flux density of the neutron energy group of the first target block at the first moment, and the volume of the first target block are multiplied to obtain the seventh target value of the neutron energy group.

[0175] In step S803 of some embodiments, the sum of the fifth target values ​​of each neutron energy group of the first target block at the first time moment is calculated to obtain the fission power of the first target block.

[0176] In some embodiments, the core transient simulation results include the core power of the first target block at the first moment, and after step S803, the results may include, but are not limited to, the following:

[0177] Obtain the decay heat of the first target node at the first time point, and based on the decay heat and fission power of the first target node at the first time point, obtain the nuclear power of the first target node at the first time point.

[0178] This embodiment provides a formula for calculating the decay heat of a segment, as shown in the following formula (21):

[0179]

[0180] Among them, P d P represents the decay heat of the nodule, ij represents the number of the fission products of the nodule, and P represents the decay heat of the nodule. dfij P represents the decay heat of the fission product ij of the nodal block under the condition of considering neutron capture. U express 239The decay heat contribution of U, P Np express 239 The decay heat contribution of Np, t n+1 This indicates the current moment, or the first moment.

[0181] This embodiment also provides a formula for calculating the nuclear power of a node, as shown in the following formula (22):

[0182]

[0183] Among them, P ijk This represents the core power of the ijk sub-block. This represents the fission power of the ijk node. This represents the decay heat of the ijk node.

[0184] In step S802 of some embodiments, the decay heat of the first target block at the first moment is obtained based on the calculation formula of the decay heat of the block, and the sum of the decay heat and fission power of the first target block at the first moment is calculated to obtain the nuclear power of the first target block at the first moment.

[0185] In some embodiments, the transient simulation results of the reactor core may also include the total fission power, total decay heat, or total nuclear power of the reactor core at the first moment.

[0186] The total nuclear power of the reactor core at the first moment is obtained by summing the nuclear power of each node in the core at the first moment. The total fission power P f It can be calculated using the following formula (23):

[0187]

[0188] The total decay heat of the core at the first moment is obtained by summing the decay heat of each node in the core. The total decay heat P is... d It can be calculated using the following formula (24):

[0189]

[0190] The total nuclear power of the reactor core at the first moment is obtained by summing the nuclear power of each node in the core. The total nuclear power P can be calculated using the following formula (25):

[0191] P(t n+1 )=∑ i＝1,j＝1,k＝1 P ijk (t n+1 (25)

[0192] In some embodiments, the core transient simulation results include the input values ​​of the external neutron detector, which simulate the result of core neutron leakage to the external neutron detector. Step S103 may include, but is not limited to, steps S901 to S903:

[0193] Step S901: For each neutron energy group of the measurement node measured by the external neutron detector, based on the target thermal parameters at the first moment and the neutron integral balance equation of the reactor core, determine the first average neutron flux density of the neutron energy group of the measurement node at the first moment.

[0194] Step S902: For each measurement node measured by the external neutron detector, the eighth target value of the first partition is determined based on the hot group neutron flux density of the measurement node corresponding to the first partition of the external neutron detector at the first time, the sensitivity coefficient of the external neutron detector to the measurement node, and the length of the first partition; the first partition is any partition among the partitions of the external neutron detector; the ninth target value of the measurement node is determined based on the eighth target value of each partition.

[0195] Step S903: Determine the input value of the external neutron detector at the first moment based on the ninth target value of each measurement block.

[0196] This embodiment also provides a formula for calculating the input value of an off-pile neutron detector, as shown in the following formula (26):

[0197]

[0198] In the formula, A dect Indicates the input value of the neutron detector; I represents the number of partitions in the neutron detector; N represents the number of measurable segments in the neutron detector; s n This represents the sensitivity coefficient of the neutron detector to the measurement node n; The thermal group average neutron flux density represents the measurement node n corresponding to the i-th partition of the detector; i t represents the length of the i-th partition of the neutron detector. n+1 This indicates the current moment, or the first moment.

[0199] In this embodiment, This is denoted as the eighth target value. This is denoted as the ninth target value.

[0200] The aforementioned measurement nodes are those that can be measured by the off-pile neutron detector, and the aforementioned first partition is any partition among the various partitions of the neutron detector.

[0201] In step S901 of some embodiments, the first average neutron flux density of each neutron energy group of the measurement block at the first time moment is calculated by the method of steps S501-S503 described above.

[0202] In step S902 of some embodiments, for each measurement node measured by the external neutron detector, the product of the hot group neutron flux density of the measurement node corresponding to the first partition of the external neutron detector at the first time moment, the sensitivity coefficient of the external neutron detector to the measurement node, and the length of the first partition is calculated to determine the eighth target value of the first partition. The sum of the eighth target values ​​for each partition is then calculated to obtain the ninth target value for the measurement block.

[0203] In step S903 of some embodiments, the sum of the ninth target values ​​of each measurement node is calculated to obtain the input value A of the external neutron detector at the first moment. dect .

[0204] In this embodiment of the application, the average neutron flux density can be obtained more accurately through the neutron integral balance equation. Subsequently, the transient simulation results of each core can be calculated using the average neutron flux density, thereby obtaining more accurate core transient simulation results and improving the effectiveness of core transient simulation.

[0205] It should be noted that the various optional implementation methods described in the embodiments of this application can be combined with each other or implemented individually without conflict, and the embodiments of this application do not limit this.

[0206] For ease of understanding, a specific embodiment will be used as an example:

[0207] Step 1: The core transient model receives the reset command from the simulation platform, performs initial loading of the core transient model, and establishes a communication interface with the simulation platform.

[0208] Step 2: Read the model input card for the core transient model, complete the initialization of the core transient model, and establish the core simulation model of the nuclear reactor.

[0209] Step 3: The core transient model receives control commands from the simulation platform. If a freeze command is received, step 11 is executed directly; if a step or run command is received, transient calculations are performed at the current moment.

[0210] Step 4: The core transient model receives the thermal parameters of the current moment from the thermal-hydraulic model through the simulation platform interface, and updates the data of each section at the current moment according to equations (11), (12), and (13).

[0211] Step 5: By defining the dynamic frequency expression, the time term in the dynamic frequency expression is differentially quantified using the implicit difference method to obtain the difference expression of the dynamic frequency. The difference expression is then substituted into the neutron kinematic equation to obtain the transient neutron diffusion equation. The transient neutron diffusion equation is spatially discretized using the semi-analytical nodal method, and volume integration is performed within the nodal. A coupling correction factor is introduced to obtain the expression of the neutron integral equilibrium equation for the entire core (6).

[0212] Given the average neutron flux density of the block at the previous time step, the coupling correction factor at the current time is calculated. Equation (6) above is then used to obtain the average neutron flux density of the neutron energy group at the current time. Specifically, the average neutron source term within the block is calculated according to equation (7), and the net neutron flux and its correction flux are solved using the transverse integration method and the residual weight method. Then, the coupling correction factor at the current time is solved. Specifically, the coupling correction factor of the neutron energy group at the current time in each coordinate direction is calculated according to equations (8), (9), and (10).

[0213] Repeat the above process until the neutron flux density calculation results satisfy the iterative convergence.

[0214] The density of delayed neutron precursor nuclei is calculated based on the above formula (14).

[0215] Step 6: Based on the neutron flux density at the current moment in Step 5, calculate the concentration of fission products of each block at the current moment according to the above formulas (16), (17), (18) and (19).

[0216] Step 7: Calculate the fission power, decay heat and nuclear power of each segment according to the above formulas (20), (21) and (22).

[0217] Calculate the total fission power, decay heat, and nuclear power of the reactor core according to the above formulas (23), (24), and (25).

[0218] Step 8: Calculate the input value of the external neutron detector at the current moment according to the above formula (26) to simulate the result of core neutron leakage to the external neutron detector.

[0219] Step 9: Synchronously update and modify parameters such as thermal parameters and control commands online through the simulation platform.

[0220] Step 10: Determine whether a new control command has been received from the simulation platform. If a storage command is received, output the calculation results to the IC file; if a backtracking command is received, execute the run backtracking function.

[0221] Step 11: Repeat steps 3 to 10 according to the control instructions of the simulation platform until the end of the run instruction is received.

[0222] Figure 2 This is a schematic diagram of the structure of the transient core simulation device for a nuclear reactor provided in this application embodiment. Please refer to [link / reference]. Figure 2 This application embodiment also provides a nuclear reactor core transient simulation device 800, applied to electronic equipment. The electronic equipment is equipped with a simulation platform, on which a thermal-hydraulic model and a core simulation model of the nuclear reactor are built. This platform can realize the above-mentioned nuclear reactor core transient simulation method. The nuclear reactor core transient simulation device 800 includes:

[0223] The first sending module 801 is used to send the target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal-hydraulic model.

[0224] The second sending module 802 is used to send transient calculation commands for the nuclear reactor and the target thermal parameters to the core simulation model through the simulation platform;

[0225] The transient simulation module 803 is used to respond to the transient calculation command through the core simulation model, and determine the core transient simulation result of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core; wherein, the neutron integral equilibrium equation is spatially discretized to determine the transient neutron diffusion equation through a nonlinear semi-analytical nodal method, the transient neutron diffusion equation is calculated through an implicit difference method, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal.

[0226] The specific implementation of the core transient simulation device for this nuclear reactor can be found in the specific embodiments of the core transient simulation method for the nuclear reactor described above, and will not be repeated here.

[0227] This application also provides an electronic device, which includes a memory and a processor. The memory stores a computer program, and the processor executes the computer program to implement the aforementioned nuclear reactor core transient simulation method. This electronic device can be any smart terminal, including tablet computers, in-vehicle computers, etc.

[0228] Please see Figure 3 , Figure 3 The hardware structure of an electronic device according to another embodiment is illustrated. The electronic device includes:

[0229] The processor 101 can be implemented using a general-purpose CPU (Central Processing Unit), microprocessor, application-specific integrated circuit (ASIC), or one or more integrated circuits, and is used to execute relevant programs to implement the technical solutions provided in the embodiments of this application.

[0230] The memory 102 can be implemented as a read-only memory (ROM), static storage device, dynamic storage device, or random access memory (RAM). The memory 102 can store the operating system and other applications. When the technical solutions provided in the embodiments of this specification are implemented through software or firmware, the relevant program code is stored in the memory 102 and is called and executed by the processor 101 to execute the core transient simulation method of the nuclear reactor of this application embodiment.

[0231] Input / output interface 103 is used to implement information input and output;

[0232] The communication interface 104 is used to enable communication and interaction between this device and other devices. Communication can be achieved through wired means (such as USB, network cable, etc.) or wireless means (such as mobile network, WIFI, Bluetooth, etc.).

[0233] Bus 105 transmits information between various components of the device (e.g., processor 101, memory 102, input / output interface 103, and communication interface 104);

[0234] The processor 101, memory 102, input / output interface 103 and communication interface 104 are connected to each other within the device via bus 105.

[0235] This application also provides a computer-readable storage medium storing a computer program that, when executed by a processor, implements the above-described nuclear reactor core transient simulation method.

[0236] Memory, as a non-transitory computer-readable storage medium, can be used to store non-transitory software programs and non-transitory computer-executable programs. Furthermore, memory may include high-speed random access memory, and may also include non-transitory memory, such as at least one disk storage device, flash memory device, or other non-transitory solid-state storage device. In some embodiments, memory may optionally include memory remotely located relative to the processor, and these remote memories can be connected to the processor via a network. Examples of such networks include, but are not limited to, the Internet, intranets, local area networks, mobile communication networks, and combinations thereof.

[0237] The embodiments described in this application are for the purpose of more clearly illustrating the technical solutions of the embodiments of this application, and do not constitute a limitation on the technical solutions provided by the embodiments of this application. As those skilled in the art will know, with the evolution of technology and the emergence of new application scenarios, the technical solutions provided by the embodiments of this application are also applicable to similar technical problems.

[0238] Those skilled in the art will understand that the technical solutions shown in the figures do not constitute a limitation on the embodiments of this application, and may include more or fewer steps than shown, or combine certain steps, or different steps.

[0239] The device embodiments described above are merely illustrative. The units described as separate components may or may not be physically separate; that is, they may be located in one place or distributed across multiple network units. Some or all of the modules can be selected to achieve the purpose of this embodiment according to actual needs.

[0240] Those skilled in the art will understand that all or some of the steps in the methods disclosed above, as well as the functional modules / units in the systems and devices, can be implemented as software, firmware, hardware, or suitable combinations thereof.

[0241] The terms “first,” “second,” “third,” “fourth,” etc. (if present) in the specification and accompanying drawings of this application are used to distinguish similar objects and are not necessarily used to describe a specific order or sequence. It should be understood that such data can be interchanged where appropriate so that the embodiments of this application described herein can be implemented in orders other than those illustrated or described herein. Furthermore, the terms “comprising” and “having,” and any variations thereof, are intended to cover non-exclusive inclusion; for example, a process, method, system, product, or apparatus that comprises a series of steps or units is not necessarily limited to those steps or units explicitly listed, but may include other steps or units not explicitly listed or inherent to such processes, methods, products, or apparatus.

[0242] It should be understood that in this application, "at least one (item)" means one or more, and "more than" means two or more. "And / or" is used to describe the relationship between related objects, indicating that three relationships can exist. For example, "A and / or B" can represent three cases: only A exists, only B exists, and both A and B exist simultaneously, where A and B can be singular or plural. The character " / " generally indicates that the preceding and following related objects are in an "or" relationship. "At least one (item) of the following" or similar expressions refer to any combination of these items, including any combination of single or plural items. For example, at least one (item) of a, b, or c can represent: a, b, c, "a and b", "a and c", "b and c", or "a and b and c", where a, b, and c can be single or multiple.

[0243] In the several embodiments provided in this application, it should be understood that the disclosed apparatus and methods can be implemented in other ways. For example, the apparatus embodiments described above are merely illustrative; for instance, the division of the units described above is only a logical functional division, and in actual implementation, there may be other division methods. For example, multiple units or components may be combined or integrated into another system, or some features may be ignored or not executed. Furthermore, the coupling or direct coupling or communication connection shown or discussed may be through some interfaces; the indirect coupling or communication connection between apparatuses or units may be electrical, mechanical, or other forms.

[0244] The units described above as separate components may or may not be physically separate. The components shown as units may or may not be physical units; that is, they may be located in one place or distributed across multiple network units. Some or all of the units can be selected to achieve the purpose of this embodiment according to actual needs.

[0245] Furthermore, the functional units in the various embodiments of this application can be integrated into one processing unit, or each unit can exist physically separately, or two or more units can be integrated into one unit. The integrated unit can be implemented in hardware or as a software functional unit.

[0246] If the integrated unit is implemented as a software functional unit and sold or used as an independent product, it can be stored in a computer-readable storage medium. Based on this understanding, the technical solution of this application, in essence, or the part that contributes to the prior art, or all or part of the technical solution, can be embodied in the form of a software product. This computer software product is stored in a storage medium and includes multiple instructions to cause a computer device (which may be a personal computer, server, or network device, etc.) to execute all or part of the steps of the methods of the various embodiments of this application. The aforementioned storage medium includes various media capable of storing programs, such as USB flash drives, portable hard drives, read-only memory (ROM), random access memory (RAM), magnetic disks, or optical disks.

[0247] The preferred embodiments of the present application have been described above with reference to the accompanying drawings, but this does not limit the scope of the claims of the present application. Any modifications, equivalent substitutions, and improvements made by those skilled in the art without departing from the scope and substance of the embodiments of the present application shall be within the scope of the claims of the present application.

Claims

1. A method for transient core simulation of a nuclear reactor, characterized in that, Applied to electronic devices, the electronic devices are equipped with a simulation platform, and the simulation platform is used to build the thermal-hydraulic model and the core simulation model of the nuclear reactor; The target thermal parameters of the nuclear reactor at the first moment are sent to the simulation platform through the thermal-hydraulic model. The simulation platform sends transient calculation commands for the nuclear reactor and the target thermal parameters to the core simulation model. The core simulation model responds to the transient calculation command and, based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core, determines the core transient simulation result of the nuclear reactor at the first moment. Specifically, the neutron integral equilibrium equation is spatially discretized using a nonlinear semi-analytical nodal method to determine the transient neutron diffusion equation, which is calculated using an implicit difference method. Furthermore, the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal. The transient simulation results of the reactor core include the slow-emitting neutron precursor nuclear density of the first target block at the first moment, wherein the first target block is any block in the reactor core of the nuclear reactor, and the first target block includes multiple neutron energy groups; The determination of the core transient simulation results of the nuclear reactor at the first moment, based on the target thermal parameters and the neutron integral equilibrium equations of the reactor core, includes: For each neutron energy group of the first target block, based on the target thermal parameters at the second time and the neutron integral balance equation of the reactor core, the first average neutron flux density of the neutron energy group of the first target block at the second time is determined, where the second time is the time before the first time. For each neutron energy group of the first target segment, the second cross-sectional data of the macroscopic fission cross-section of the neutron energy group at the second time is determined according to the target thermal parameters at the second time; and the first target value of the neutron energy group is determined according to the second cross-sectional data of the neutron energy group and the first average neutron flux density of the neutron energy group of the first target segment at the second time. The delayed neutron precursor nuclear density of the first target block at the first time step is determined based on the first target value of each of the neutron energy groups and the delayed neutron precursor nuclear density of the first target block at the second time step.

2. The method according to claim 1, characterized in that, Before the neutron integral equilibrium equations based on the target thermal parameters and the reactor core, the method further includes: The time derivative of the neutron flux density is used as the first parameter; The ratio of the first parameter to the neutron flux density is used as the first dynamic frequency. The time derivative of the delayed neutron precursor nuclear density is used as the second parameter; The ratio of the second parameter to the density of the neutron precursor nucleus is used as the second dynamic frequency. The first dynamic frequency and the second dynamic frequency are discretized by implicit difference method to obtain the difference representation of the first dynamic frequency and the difference representation of the second dynamic frequency. The transient neutron diffusion equation is obtained based on the differential representation of the first dynamic frequency and the differential representation of the second dynamic frequency. The transient neutron diffusion equation is spatially discretized to obtain the neutron integral equilibrium equation.

3. The method according to claim 1, characterized in that, The neutron integral equilibrium equation is used to characterize the relationship between the first average neutron flux density of the neutron energy group of the first block at the first time, the second average neutron flux density of the neutron energy group of the second block at the first time, and the average neutron source term within the block of the neutron energy group of the first block; wherein, the second block is the adjacent block of the first block in each coordinate direction. Before the neutron integral equilibrium equations based on the target thermal parameters and the reactor core, the method further includes: For each second segment, a first coefficient corresponding to the second average neutron flux density of the second segment is determined based on the target parameters of the first segment and the second segment in the target coordinate direction corresponding to the second segment, and the coupling correction factor of the first segment in the target coordinate direction; wherein, the target parameters in the target coordinate direction include the first size and diffusion coefficient of the first segment in the target coordinate direction, and the second size of the second segment in the target coordinate direction; The second coefficient corresponding to the first average neutron flux density of the first block is determined based on the first coefficient corresponding to the second average neutron flux density of each second block and the macroscopic neutron removal cross section of the neutron energy group. The neutron integral balance equation is determined based on each of the first coefficient, the second coefficient, the first average neutron flux density at the first time, the second average neutron flux density at the first time, and the average neutron source term within the block.

4. The method according to claim 3, characterized in that, Before determining the first coefficient corresponding to the second average neutron flux density of the second segment based on the target parameters of the first segment and the second segment in the target coordinate direction corresponding to the second segment, and the coupling correction factor of the first segment in the target coordinate direction, the method further includes: Based on the diffusion coefficient of the first segment in the target coordinate direction, and the characteristic parameters of the first segment and the second segment corresponding to the target coordinate direction, the net neutron differential flux of the neutron energy group on the surface of the first segment in the target coordinate direction is determined; wherein, the characteristic parameters include average neutron flux density and size; Based on the net neutron differential current of the neutron energy group on the surface of the first segment in the target coordinate direction and the net neutron current of the neutron energy group on the surface of the segment in the target coordinate direction, determine the net neutron correction current of the neutron energy group on the surface of the first segment in the target coordinate direction. Based on the net neutron correction flow and the characteristic parameters of the first block and the second block corresponding to the target coordinate direction, the coupling correction factor of the first block in the target coordinate direction is determined.

5. The method according to claim 3, characterized in that, The transient simulation results of the reactor core include the first average neutron flux density of the first neutron energy group of the first target block at the first moment, wherein the first target block is any block in the reactor core of the nuclear reactor, the first target block includes multiple neutron energy groups, and the first neutron energy group is any neutron energy group of the first target block; The determination of the core transient simulation results of the nuclear reactor at the first moment, based on the target thermal parameters and the neutron integral equilibrium equations of the reactor core, includes: Based on the target thermal parameters at the first moment, determine the first cross-sectional data of the macroscopic neutron removal cross section of the first neutron energy group at the first moment; Obtain the coupling correction factor of the first neutron energy group in each coordinate direction at the first time point, the target parameter, and the average neutron source term within the block; The first cross-sectional data of the first neutron energy group at the first time point, the coupling correction factor of the first neutron energy group in each coordinate direction at the first time point, the target parameter, and the average neutron source term within the block are input into the neutron integral balance equation to obtain the first average neutron flux density at the first time point.

6. The method according to claim 1, characterized in that, The transient simulation results of the reactor core include the concentration of fission products of the first target block at the first time point, wherein the first target block is any block in the reactor core of the nuclear reactor, the first target block includes multiple neutron energy groups, and the fission products include at least one of fission product 135I, fission product 135Xe, fission product 149Pm, and fission product 149Sm. The determination of the core transient simulation results of the nuclear reactor at the first moment, based on the target thermal parameters and the neutron integral equilibrium equations of the reactor core, includes: The concentration of fission products of the first target block is obtained at the second time point, which is the time point preceding the first time point; For each neutron energy group of the first target block, based on the target thermal parameters at the first moment and the neutron integral balance equation of the nuclear reactor, the first average neutron flux density of the neutron energy group of the first target block at the first moment is determined. For each neutron energy group of the first target block, based on the target thermal parameters at the first moment, the cross-sectional data of the macroscopic absorption cross section of the fission products of the neutron energy group are determined, and based on the cross-sectional data of the macroscopic absorption cross section of the fission products of the neutron energy group and the first average neutron flux density of the neutron energy group at the first moment, the second target value of the fission products of the neutron energy group at the first moment is determined. The concentration change of the fission products from the second time point to the first time point is obtained, wherein the concentration change is determined by target information, which includes relevant parameters of the fission products and the concentration of the fission products in the first target block at the second time point. The relevant parameters of the fission products include the fission yield and decay constant of the fission products. If the fission products are 135I or 149Pm, the target information also includes a second target value of the fission products of each neutron energy group in the first target block at the first time point. If the fission products are 135Xe, the target information also includes a third target value, a fourth target value, and the second target value of the fission products of each neutron energy group in the first target block at the first time point. The third target value is determined based on the... The decay constant of 135I and the concentration of 135I at the first time point are used to determine the fourth target value. The fourth target value is determined based on the cross-sectional data of the microscopic absorption cross-section of 135Xe of each neutron energy group of the first target block at the first time point, and the first average neutron flux density of each neutron energy group at the first time point. If the fission product is 149Sm, the target information also includes a fifth target value and a sixth target value. The fifth target value is determined based on the decay constant of 149Pm and the concentration of 149Pm at the first time point. The sixth target value is determined based on the cross-sectional data of the microscopic absorption cross-section of 149Sm of each neutron energy group of the first target block at the first time point, and the first average neutron flux density of each neutron energy group at the first time point. The concentration of fission products of the first target block at the first time point is determined based on the concentration of fission products of the first target block at the second time point and the change in concentration of the fission products from the second time point to the first time point.

7. The method according to claim 1, characterized in that, The core transient simulation results include the fission power of the first target block, wherein the first target block is any block in the core of the nuclear reactor, and the first target block includes multiple neutron energy groups; The determination of the core transient simulation results of the nuclear reactor at the first moment, based on the target thermal parameters and the neutron integral equilibrium equations of the reactor core, includes: For each neutron energy group of the first target segment, based on the target thermal parameters at the first moment and the neutron integral balance equation of the reactor core, the first average neutron flux density of the neutron energy group of the first target segment at the first moment is determined. For each neutron energy group of the first target block, the cross-sectional data of the macroscopic energy generation section of the neutron energy group is determined according to the target thermal parameters at the first moment; and the seventh target value of the neutron energy group is determined according to the cross-sectional data of the macroscopic energy generation section of the neutron energy group, the first average neutron flux density of the neutron energy group of the first target block at the first moment, and the volume of the first target block. The fission power of the first target block at the first moment is determined based on the seventh target value of each neutron energy group of the first target block. The transient simulation results of the reactor core include the nuclear power of the first target block at the first moment. After determining the fission power of the first target block based on the seventh target value of each of the neutron energy groups of the first target block, the method further includes: Obtain the decay heat of the first target node at the first time point, and based on the decay heat and fission power of the first target node at the first time point, obtain the nuclear power of the first target node at the first time point.

8. The method according to claim 1, characterized in that, The core transient simulation results include the input values ​​of the external neutron detector. The determination of the core transient simulation results of the nuclear reactor at the first moment, based on the target thermal parameters and the neutron integral equilibrium equation of the reactor core, includes: For each neutron energy group of the measurement node measured by the off-pile neutron detector, the first average neutron flux density of the neutron energy group of the measurement node at the first moment is determined based on the target thermal parameters at the first moment and the neutron integral balance equation of the reactor core. For each measurement node measured by the external neutron detector, the eighth target value of the first partition is determined based on the hot group neutron flux density of the measurement node corresponding to the first partition of the external neutron detector at the first time, the sensitivity coefficient of the external neutron detector to the measurement node, and the length of the first partition; the first partition is any partition among the partitions of the external neutron detector; the ninth target value of the measurement node is determined based on the eighth target value of each partition. The input value of the external neutron detector at the first moment is determined based on the ninth target value of each measurement node.

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