Nuclear reactor core steady-state simulation methods and related devices

Abstract

This application provides a method and related apparatus for core steady-state 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 steady-state calculation instructions and the target thermal parameters to the core simulation model. Responding to the steady-state calculation instructions, the core simulation model determines the core steady-state 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 steady-state model to be instructed to perform core steady-state simulation via steady-state calculation instructions. The core steady-state model intelligently performs core steady-state calculations based on the target thermal parameters and obtains the calculation results, avoiding manual operation and improving computational efficiency.

Description

Technical Field

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

[0002] Currently, for core steady-state calculations of nuclear reactors, the common practice is to manually operate the core steady-state analysis software function after obtaining the thermal parameters from the nuclear reactor simulation on the simulation platform. This results in slow core steady-state analysis efficiency and the inability to obtain core steady-state simulation data in a timely manner. Summary of the Invention

[0003] The main objective of this application is to propose a core steady-state simulation method and related apparatus for nuclear reactors, aiming to improve the efficiency of core steady-state simulation.

[0004] To achieve the above objectives, a first aspect of this application proposes a core steady-state 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:

[0005] The target thermal parameters of the nuclear reactor at the first moment are sent to the simulation platform through the thermal-hydraulic model.

[0006] The simulation platform sends steady-state calculation commands for the nuclear reactor and the target thermal parameters to the core simulation model.

[0007] The core simulation model responds to the steady-state calculation command and determines the core steady-state 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. The neutron integral equilibrium equation is spatially discretized to determine the neutron steady-state diffusion equation through a nonlinear semi-analytical nodal method, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal.

[0008] In some embodiments, the above-described 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, the second average neutron flux density of the neutron energy group of the second block, 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;

[0009] Before the neutron integral equilibrium equations based on the target thermal parameters and the reactor core, the method further includes:

[0010] 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 correction coupling 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;

[0011] 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.

[0012] The neutron integral balance equation is determined based on each of the first coefficient, the second coefficient, the first average neutron flux density, the second average neutron flux density, and the average neutron source term within the block.

[0013] In some embodiments, 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 correction coupling factor of the first segment in the target coordinate direction, the method further includes:

[0014] 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;

[0015] 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.

[0016] 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 correction coupling factor of the first block in the target coordinate direction is determined.

[0017] In some embodiments, the above-mentioned core steady-state simulation results include the first average neutron flux density of the first neutron energy group of the first target node, wherein the first target node is any node in the core of the nuclear reactor, and the first neutron energy group is any neutron energy group of the first target node;

[0018] The determination of the core steady-state 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:

[0019] Based on the target thermal parameters, determine the first cross-sectional data of the macroscopic neutron removal cross-section of the first neutron energy group;

[0020] Obtain the corrected coupling factor of the first neutron energy group in each coordinate direction, the target parameter, and the average neutron source term within the block;

[0021] The first cross-sectional data of the first neutron energy group, the corrected coupling factor of the first neutron energy group in each coordinate direction, 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.

[0022] In some embodiments, the above-mentioned core steady-state simulation results include the steady-state value of the delayed neutron precursor nuclear density of the first target node, wherein the first target node is any node in the core of the nuclear reactor;

[0023] The determination of the core steady-state 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:

[0024] For each neutron energy group of the first target segment, the first average neutron flux density of the neutron energy group of the first target segment is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0025] For each neutron energy group of the first target block, the second cross-sectional data of the macroscopic fission cross section of the neutron energy group is determined according to the target thermal parameters; 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.

[0026] Based on the first target value of each of the neutron energy groups, determine the steady-state value of the slow-emission neutron precursor nuclear density of the first target block.

[0027] In some embodiments, the above-mentioned core steady-state simulation results include the steady-state concentration values ​​of the fission products of the first target node, wherein the first target node is any node in the core of the nuclear reactor, and the fission products include at least one of fission product 135I, fission product 135Xe, fission product 149Pm, and fission product 149Sm.

[0028] The determination of the core steady-state 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, further includes:

[0029] For each neutron energy group of the first target segment, the first average neutron flux density of the neutron energy group of the first target segment is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0030] For each neutron energy group of the first target block, the third cross-sectional data of the macroscopic absorption cross-section of the fission products of the neutron energy group is determined according to the target thermal parameters, and the second target value of the fission products of the neutron energy group is determined according to the third cross-sectional data of the fission products and the first average neutron flux density of the neutron energy group.

[0031] Obtain the fourth coefficient corresponding to the fission product; wherein, if the fission product is fission product 135I or fission product 149Pm, the fourth coefficient is determined by the relevant parameters of the fission product, the relevant parameters including fission yield and decay constant; if the fission product is fission product 135Xe, the fourth coefficient is determined by the fission yield of fission product 135I, the relevant parameters of fission product 135Xe, and a third target value, the third target value being based on the neutron energy groups of the first target block in the first... The cross-sectional data of the microscopic absorption cross-section of the fission product 135Xe at the first time and the first average neutron flux density of each neutron energy group are used to determine the fourth coefficient. If the fission product is fission product 149Sm, the fourth coefficient is determined by the fission yield of fission product 149Sm and the fourth target value of fission product 149Sm. The fourth target value is determined based on the cross-sectional data of the microscopic absorption cross-section of the fission product 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.

[0032] The steady-state concentration of the fission products of the first target node is determined based on the second target value of the fission products of each of the neutron energy groups and the fourth coefficient.

[0033] In some embodiments, the above-mentioned core steady-state simulation results include the fission power of the first target segment, wherein the first target segment is any segment in the core of the nuclear reactor;

[0034] The determination of the core steady-state 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, further includes:

[0035] For each neutron energy group of the first target segment, the first average neutron flux density of the neutron energy group of the first target segment is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0036] For each neutron energy group of the first target block, the fourth cross-sectional data of the macroscopic energy generation cross-section of the neutron energy group is determined according to the target thermal parameters; and the fifth target value of the neutron energy group is determined according to the fourth cross-sectional data of the neutron energy group, the first average neutron flux density and the volume of the first target block.

[0037] The fission power of the first target block is determined based on the fifth target value of each of the neutron energy groups of the first target block.

[0038] The core steady-state simulation results include the nuclear power of the first target node. After determining the fission power of the first target node based on the fifth target value of each of the neutron energy groups of the first target node, the process further includes:

[0039] Obtain the decay heat of the first target node, and based on the decay heat and fission power of the first target node, obtain the nuclear power of the first target node.

[0040] In some embodiments, the above-mentioned core steady-state simulation results include input values ​​from at least one section of the external neutron detector. The determination of the core steady-state 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, further includes:

[0041] For the measurement nodes measured by at least one partition of the off-pile neutron detector, the first average neutron flux density of each neutron energy group of the measurement node is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0042] For each of the at least one partition, a sixth target value for the target measurement node is determined based on the hot-group average neutron flux density of the target measurement node corresponding to the partition, the sensitivity coefficient of the off-pile neutron detector to the target measurement node, and the length of the partition; wherein, the target measurement node is any node among the measurement nodes corresponding to the partition;

[0043] For each partition of the at least one partition, the input value of the partition is determined based on the sixth target value of each measurement block corresponding to the partition.

[0044] To achieve the above objectives, a second aspect of this application provides a core steady-state simulation device for a nuclear reactor, the device comprising:

[0045] A thermal-hydraulic model is used to send the target thermal parameters of the nuclear reactor at the first moment to the simulation platform.

[0046] Simulation platform: used to send steady-state calculation commands for the nuclear reactor and the target thermal parameters to the core simulation model.

[0047] A core simulation model, in response to the steady-state calculation command, determines the core steady-state 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. The neutron integral equilibrium equation is spatially discretized using a nonlinear semi-analytical nodal method to determine the neutron steady-state diffusion equation, and the intra-nodal average neutron source term in the neutron integral equilibrium equation is determined by volume integration within the nodal.

[0048] 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.

[0049] 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.

[0050] This application proposes a core steady-state simulation method and related apparatus for nuclear reactors. The method is applied to electronic equipment equipped with a simulation platform. The simulation platform is used to build a thermal-hydraulic model and a core simulation model of the nuclear reactor. The thermal-hydraulic model sends the target thermal parameters of the nuclear reactor at the first moment to the simulation platform. The simulation platform sends steady-state calculation instructions and the target thermal parameters to the core simulation model. The core simulation model responds to the steady-state calculation instructions and determines the core steady-state 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. Among them, the neutron integral equilibrium equation is spatially discretized to determine the neutron steady-state diffusion equation through a nonlinear semi-analytical nodal method, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal. Thus, this invention provides a core steady-state model that can be instructed to perform core steady-state simulations via steady-state calculation commands. The core steady-state model, responding to these commands, automatically and intelligently performs core steady-state calculations based on target thermal parameters, obtaining the results and avoiding manual operation, thereby improving computational efficiency. Furthermore, since the nonlinear semi-analytical nodal method has high accuracy and computational efficiency, spatial discretization of the neutron steady-state diffusion equation using this method yields a more accurate and efficient neutron integral equilibrium equation, which is beneficial for obtaining more effective core steady-state simulation results. Attached Figure Description

[0051] Figure 1 This is a flowchart of a core steady-state simulation method for a nuclear reactor provided in an embodiment of this application;

[0052] Figure 2 yes Figure 1 A flowchart of the steps preceding step S103;

[0053] Figure 3 yes Figure 2 A flowchart of the steps preceding step S201;

[0054] Figure 4 yes Figure 1 The flowchart of step S103 in the process;

[0055] Figure 5 yes Figure 1 The flowchart of step S103 in the process;

[0056] Figure 6 yes Figure 1 The flowchart of step S103 in the process;

[0057] Figure 7 yes Figure 1 The flowchart of step S103 in the process;

[0058] Figure 8 yes Figure 1 The flowchart of step S103 in the process;

[0059] Figure 9 This is a schematic diagram of the structure of the nuclear reactor core steady-state simulation device provided in the embodiments of this application;

[0060] Figure 10 This is a schematic diagram of the hardware structure of the electronic device provided in the embodiments of this application. Detailed Implementation

[0061] 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.

[0062] 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.

[0063] 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.

[0064] This application provides a method, apparatus, electronic device, and storage medium for core steady-state simulation of a nuclear reactor, aiming to improve the efficiency of core steady-state simulation.

[0065] The core steady-state simulation method, apparatus, electronic equipment, and storage medium for nuclear reactors provided in this application are specifically illustrated through the following embodiments. First, the core steady-state simulation method for nuclear reactors in this application is described.

[0066] The core steady-state simulation method for nuclear reactors provided in this application relates to the field of artificial intelligence technology. This core steady-state simulation 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.

[0067] Figure 1This is an optional flowchart of a core steady-state simulation method for a nuclear reactor provided in this application embodiment. The method is applied to an electronic device equipped with a simulation platform. The simulation platform is used to simulate the 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 steady-state simulation of the nuclear reactor.

[0068] Figure 1 The method may include, but is not limited to, steps S101 to S103.

[0069] Step S101: Send the target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal-hydraulic model.

[0070] Step S102: Send steady-state calculation commands and target thermal parameters for the nuclear reactor to the core simulation model through the simulation platform.

[0071] Step S103: Responding to the steady-state calculation command through the core simulation model, the core steady-state simulation results of the nuclear reactor at the first moment are determined based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core.

[0072] The aforementioned steady-state calculation command is used to instruct the core simulation model to perform steady-state simulation of the nuclear reactor core.

[0073] In step S101 of some embodiments, the thermal-hydraulic model automatically feeds back thermal parameters to the simulation platform at each preset calculation cycle (time step).

[0074] In step S102 of some embodiments, when the core simulation model needs to run a core steady-state 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. The operation command is used to instruct the simulation platform to send the steady-state calculation command 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 the steady-state calculation command for the nuclear reactor and the target thermal parameters of the nuclear reactor at the first moment to the core simulation model.

[0075] The aforementioned first moment can be the latest moment (current moment), which can enable the simulation platform to send steady-state calculation instructions 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.

[0076] In some embodiments, before step S103, a core steady-state model is first established. The core steady-state model receives a reset command from the simulation platform and resets (initial loads) the core steady-state model based on the reset command, establishes a communication interface with the simulation platform, and reads the model input card. The core steady-state model is initialized based on the model input card, and a core simulation model of the nuclear reactor is established. The aforementioned model input card may include core layout, node size, or node division, etc.

[0077] 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 steady-state model to perform corresponding operations. These commands include freeze commands, run commands (steady-state calculation commands), store commands, or backtrack commands. The freeze command instructs the core steady-state model to perform no processing; the store command instructs the core steady-state model to output its calculation results to an IC file; and the backtrack command instructs the core steady-state model to execute a run backtrack function. For example, the simulation platform sends steady-state 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.

[0078] In step S103 of some embodiments, the above-mentioned neutron integral balance equation is spatially discretized to determine the neutron steady-state diffusion equation by a nonlinear semi-analytical nodal method, and the average neutron source term within the nodal in the neutron integral balance equation is determined by volume integration within the nodal.

[0079] The core simulation model receives steady-state calculation commands from the simulation platform via a communication interface. In response to the steady-state calculation commands, it calculates the core steady-state 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.

[0080] Steps S101 to S103 of this embodiment involve sending steady-state 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 steady-state simulation function in response to the commands, eliminating the need for manual core steady-state simulation operations and improving simulation efficiency. Furthermore, since the nonlinear semi-analytical nodal method has high accuracy and computational efficiency, spatial discretization of the neutron steady-state diffusion equation using this method yields a more accurate and efficient neutron integral equilibrium equation, which is beneficial for obtaining more effective core steady-state simulation results. In addition, the simulation platform can send steady-state calculation commands in real-time, instructing the core simulation model to perform core steady-state simulation based on the latest thermal parameters. The calculation speed meets real-time requirements, and the process of this core steady-state simulation method can accommodate various operating conditions, making it more widely applicable.

[0081] In some embodiments, the above-described 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, the second average neutron flux density of the neutron energy group of the second block, 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. (See also...) Figure 2 In some embodiments, the neutron integral equilibrium equation is established before step S103, and the steps before step S102 may include, but are not limited to, steps S201 to S205:

[0082] Step S201: For each second block, determine the first coefficient corresponding to the second average neutron flux density of the second block based on the target parameters of the first and second blocks in the target coordinate direction corresponding to the second block, and the correction coupling factor of the first block in the target coordinate direction.

[0083] Step S202: 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.

[0084] Step S203: Determine the neutron integral balance equation based on each first coefficient, second coefficient, first average neutron flux density, second average neutron flux density, and the average neutron source term within the block.

[0085] The aforementioned coordinate directions include the positive and negative directions of the x-axis, y-axis, or z-axis in the Cartesian coordinate system. Therefore, the second block can include the first block preceding and following the first block on the x-axis, the first block preceding and following the first block on the y-axis, and the first block preceding and following the first block on the z-axis.

[0086] In this embodiment, the above-mentioned neutron integral balance equation is used to calculate the first average neutron flux density of the neutron energy group of the first block.

[0087] 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.

[0088] 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:

[0089]

[0090] In the formula, Cf1 represents the first coefficient, h represents the node size, and D represents the diffusion coefficient. The modified coupling 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 corrected coupling factor of the neutron energy group in the target coordinate direction u+.

[0091] For the first coefficient of the second block corresponding to the negative target coordinate direction, the expression for the first coefficient is as follows:

[0092]

[0093] In the formula, Cf1 represents the first coefficient, h represents the node size, and D represents the diffusion coefficient. The modified coupling 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 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 corrected coupling factor of the neutron energy group in the target coordinate direction u-.

[0094] 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.

[0095] In one implementation, the expression for the second coefficient can be as follows:

[0096]

[0097] in, Cf2The second coefficient is represented by ijk, indicating 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.

[0098] 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, and the second product is the product of the second coefficient and the corresponding second average neutron flux density.

[0099] In one embodiment, the expression for the above-mentioned neutron integral equilibrium equation is shown in the following formula (1):

[0100]

[0101] 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. This represents the block-average neutron source term of the neutron energy group in the first block.

[0102] The formula for calculating the average neutron source term within the above-mentioned block can be shown in formula (2) below:

[0103]

[0104] Among them, Q g,ijk (x,y,z) represents the neutron source term.

[0105] Please see Figure 3In some embodiments, steps S201 may include, but are not limited to, steps S301 to S304:

[0106] Step S301: Based on the diffusion coefficient of the first segment in the target coordinate direction and the characteristic parameters of the second segment corresponding to the first segment and the target coordinate direction, determine the net neutron differential current of the neutron energy group on the surface of the first segment in the target coordinate direction.

[0107] Step S302: Based on the net neutron differential current of the neutron energy group on the block surface of the first block in the target coordinate direction and the net neutron current of the neutron energy group on the block surface in the target coordinate direction, determine the net neutron correction current of the neutron energy group on the block surface of the first block in the target coordinate direction.

[0108] Step S303: 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 correction coupling factor of the first block in the target coordinate direction.

[0109] The aforementioned characteristic parameters include average neutron flux density and size.

[0110] 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.

[0111] In step S301 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 by the following formula (3):

[0112]

[0113] 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.

[0114] In step S302 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 shown in the following formula (4):

[0115]

[0116] 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.

[0117] In step S303 of some embodiments, the expression for the corrected coupling factor of the first block in the target coordinate direction is shown in the following formula (5):

[0118]

[0119] in, This represents the correction coupling factor of the first block in the target coordinate direction.

[0120] Please see Figure 4 In some embodiments, the core steady-state simulation results include the first average neutron flux density of the first neutron energy group of the first target node, where the first target node is any node in the core of the nuclear reactor, and the first neutron energy group is any neutron energy group of the first target node. Step S103 may include, but is not limited to, steps S401 to S403:

[0121] Step S401: Determine the first cross-sectional data of the macroscopic neutron removal cross section of the first neutron energy group based on the target thermal parameters.

[0122] Step S402: Obtain the corrected coupling factor, target parameters, and average neutron source terms within the block for the first neutron energy group in each coordinate direction.

[0123] Step S403: Input the first cross-sectional data of the first neutron energy group, the corrected coupling factor of the first neutron energy group in each coordinate direction, the target parameter, and the average neutron source term within the block into the neutron integral equilibrium equation to obtain the first average neutron flux density.

[0124] 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.

[0125] The formula for calculating the cross-sectional data of the control rod before correction is shown in formula (6) below:

[0126]

[0127] 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.

[0128] Then, based on the rod position update of each group of control rods and the cross-sectional data at the first moment calculated by the above formula (6), the final cross-sectional data of the neutron energy group is obtained. Among them, the cross-sectional data calculated by the rod position update formula (6) of each group of control rods is as follows:

[0129]

[0130] 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.

[0131] 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 formulas (6) and (7), 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 first moment by formula (8) as follows:

[0132]

[0133] 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 first moment are respectively; N Xe N Sm The densities are 135Xe and 149Sm, respectively.

[0134] 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 using the above-mentioned calculation formulas (6) and (7).

[0135] 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 S401-S402 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 (6) and (7), 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.

[0136] In step S402 of some embodiments, the corrected coupling factor of the first neutron energy group in each coordinate direction can be obtained by formulas (3), (4), and (5) in steps S301-S303. Since the expression of the corrected coupling factor uses the first average neutron flux density of the neutron energy group as a parameter, the corrected coupling factor of the first neutron energy group in each coordinate direction obtained by steps S301-S303 is an expression that only uses the first average neutron flux density of the first neutron energy group as an unknown.

[0137] In step S403 of some embodiments, the first cross-sectional data of the first neutron energy group, the correction coupling factor of the first neutron energy group 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 as the unknown can be obtained. The first average neutron flux density of the first neutron energy group is obtained by solving the neutron integral balance equation.

[0138] Please see Figure 5 In some embodiments, the core steady-state simulation results include the steady-state value of the delayed neutron precursor nuclear density of the first target node, and step S103 may also include, but is not limited to, steps S501 to S503:

[0139] Step S501: For each neutron energy group of the first target block, determine the first average neutron flux density of the neutron energy group of the first target block based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0140] Step S502: 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 according to the target thermal parameters; 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.

[0141] Step S503: Determine the steady-state value of the slow-emission neutron precursor nuclear density of the first target block based on the first target value of each neutron energy group.

[0142] This embodiment provides a method for calculating the nuclear density of each group of delayed neutron precursors, and the calculation formula is shown in the following formula (9):

[0143]

[0144] 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. The macroscopic fission cross section of the g′ group represents the nodal segment. β represents the average neutron flux density of the g′ group of the nodal. m λ represents the proportion of the m-th delayed neutron group in all fission neutrons. m Let represent the decay constant of the m-th slow-emitting neutron precursor nucleus.

[0145] 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.

[0146] In this embodiment, This is denoted as the first target value.

[0147] In step S501 of some embodiments, firstly, the first average neutron flux density of each neutron energy group of the first target block is calculated according to the above steps S401-S403.

[0148] In step S502 of some embodiments, the second cross-sectional data of the macroscopic fission cross-section of each neutron energy group of the first target block is calculated according to the calculation formulas (6) and (7) of the cross-sectional data. For each neutron energy group, the product between the corresponding second cross-sectional data and the first average neutron flux density is calculated to obtain the first target value of the neutron energy group.

[0149] In step S503 of some embodiments, the proportion of the m-th group of delayed neutrons in all fission neutrons and the decay constant of the m-th group of delayed neutron precursor nuclei corresponding to the first target block are obtained. According to the above formula (9), based on the proportion of the m-th group of delayed neutrons in all fission neutrons and the decay constant of the m-th group of delayed neutron precursor nuclei, and the first target value of each neutron energy group corresponding to the first target block, the steady-state value of the delayed neutron precursor nucleus density of the first target block is calculated.

[0150] Please see Figure 6In some embodiments, the core steady-state simulation results include the steady-state concentration values ​​of fission products of the first target node, wherein 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 S601 to S604:

[0151] Step S601: For each neutron energy group of the first target block, determine the first average neutron flux density of the neutron energy group of the first target block based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0152] Step S602: For each neutron energy group of the first target block, determine the third cross-sectional data of the macroscopic absorption cross-section of the fission products of the neutron energy group according to the target thermal parameters, and determine the second target value of the fission products of the neutron energy group according to the third cross-sectional data of the fission products and the first average neutron flux density of the neutron energy group.

[0153] Step S603: Obtain the fourth coefficient corresponding to the fission products.

[0154] Step S604: Determine the steady-state concentration of the fission products of the first target block based on the second target value and the fourth coefficient of the fission products of each neutron energy group.

[0155] In this embodiment, a formula for calculating the steady-state concentration of fission products in each segment is provided, wherein the formula for calculating the steady-state concentration of fission product 135I is as follows:

[0156]

[0157] in, This represents the steady-state concentration of 135I, a fission product of the nodule, γ. I λ represents the fission yield of fission product 135I. I The decay constant of the fission product 135I is represented. The macroscopic absorption cross section represents the fission product 135I of the g′ group. This represents the average neutron flux density of the nodal group g′.

[0158] The steady-state concentration of the fission product 135Xe is calculated using the following formula (11):

[0159]

[0160] in, This represents the steady-state concentration of the fission product 135Xe in the nodule, γ. I Indicates the fission yield of fission product 135I, γ Xeλ 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′ group. This represents the average neutron flux density of the nodal g′ group. The microscopic absorption cross section of the fission product 135Xe of the nodal group g′ is shown at the first moment.

[0161] The steady-state concentration of the fission product 149Pm is calculated using the following formula (12):

[0162]

[0163] in, This represents the steady-state concentration of the fission products 149Pm in the nodule, 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 nodal group g′. This represents the average neutron flux density of the nodal group g′.

[0164] The steady-state concentration of the fission product 149Sm is calculated using the following formula (13):

[0165]

[0166] in, This represents the steady-state concentration of 149Sm, a fission product of the nodule, and γ. 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 nodal group g′. This represents the average neutron flux density of the nodal g′ group. The microscopic absorption cross section of the fission product 149Sm of the nodal group g′ is shown at the first moment.

[0167] In this embodiment, the formula for calculating the steady-state concentration of each fission product is modified by removing... The portion other than that is denoted as the fourth coefficient. This is denoted as the second target value.

[0168] Therefore, in the formula for calculating the steady-state concentration of fission product 135I or fission product 149Pm, the fourth coefficient is determined by the relevant parameters of the fission product, including fission yield and decay constant. That is, the fourth coefficient in the formula for calculating the steady-state concentration of fission product 135I is... The fourth coefficient in the formula for calculating the steady-state concentration of fission product 149Pm is...

[0169] For the formula for calculating the steady-state concentration of fission product 135Xe, the fourth coefficient is composed of the fission yield of fission product 135I, relevant parameters of fission product 135Xe, and the third target value. Determined, among which, the third target value Based on the cross-sectional data of the microscopic absorption cross-section of the fission product 135Xe in each neutron energy group of the first target block at the first moment and the first average neutron flux density of each neutron energy group, the fourth coefficient of the formula for calculating the steady-state concentration of the fission product 135Xe is determined as follows:

[0170]

[0171] For the formula for calculating the steady-state concentration of fission product 149Sm, the fourth coefficient is composed of the fission yield of fission product 149Sm and the fourth target value of fission product 149Sm. Fourth target value confirmed. Based on the cross-sectional data of the microscopic absorption cross-section of the fission product 149Sm in each neutron energy group of the first target block at the first moment and the first average neutron flux density of each neutron energy group, the fourth coefficient of the formula for calculating the steady-state concentration value of the fission product 149Sm is determined as follows:

[0172] In step S601 of some embodiments, firstly, the first average neutron flux density of each neutron energy group of the first target block is calculated according to the above steps S401-S403.

[0173] In step S602 of some embodiments, the third cross-sectional data of the macroscopic absorption cross-section of the fission products of each neutron energy group of the first target block is calculated according to the calculation formulas (6), (7), and (8) of the cross-sectional data described above. For each neutron energy group, the product of the third cross-sectional data corresponding to the fission products of each neutron energy group and the first average neutron flux density is calculated to obtain the second target value of the fission products of the neutron energy group.

[0174] In step S603 of some embodiments, the fourth coefficient corresponding to the fission products is obtained according to the corresponding fourth coefficient calculation method.

[0175] In step S604 of some embodiments, the steady-state concentration of the fission products of the first target block is calculated according to the formula for calculating the steady-state concentration of fission products, based on the second target value and the fourth coefficient of each neutron energy group.

[0176] Please see Figure 7In some embodiments, the core steady-state simulation results include the fission power of the first target node, and step S102 may include, but is not limited to, steps S701 to S702:

[0177] Step S701: For each neutron energy group of the first target block, determine the first average neutron flux density of the neutron energy group of the first target block based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0178] Step S702: For each neutron energy group of the first target block, determine the fourth cross-sectional data of the macroscopic energy generation cross-section of the neutron energy group according to the target thermal parameters; and determine the fifth target value of the neutron energy group according to the fourth cross-sectional data of the neutron energy group, the first average neutron flux density and the volume of the first target block.

[0179] Step S703: Determine the fission power of the first target block based on the fifth target value of each neutron energy group of the first target block.

[0180] In this embodiment, a method for calculating the fission power of a segment is provided, and the calculation formula is shown in the following formula (14):

[0181]

[0182] in, This represents the fission power of the segment. This represents the cross section for macroscopic energy generation in the nodal group g′. V represents the average neutron flux density of the nodal g′ group. ijk This indicates the volume of the node.

[0183] In this embodiment, This is denoted as the fifth target value.

[0184] In step S701 of some embodiments, firstly, the first average neutron flux density of each neutron energy group of the first target block is calculated according to the above steps S401-S403.

[0185] In step S702 of some embodiments, the fourth cross-sectional data of the macroscopic energy generation cross-section of each neutron energy group of the first target block is calculated according to the calculation formulas (6) and (7) of the cross-sectional data described above. For each neutron energy group, the product of the fourth cross-sectional data corresponding to the neutron energy group, the first average neutron flux density, and the volume of the first target block is calculated to obtain the fifth target value of the neutron energy group.

[0186] In step S703 of some embodiments, the sum of the fifth target values ​​of each neutron energy group of the first target block is calculated to obtain the fission power of the first target block.

[0187] In some embodiments, the core steady-state simulation results include the core power of the first target node, and step S703 may include, but is not limited to, step S801:

[0188] Step S801: Obtain the decay heat of the first target node, and obtain the nuclear power of the first target node based on the decay heat and fission power of the first target node.

[0189] This embodiment provides a formula for calculating the decay heat of a segment, as shown in the following formula (15):

[0190]

[0191] Among them, P d (t) represents the decay heat of the nodule, ij represents the number of the fission products of the nodule, P dfij P represents the decay heat of the fission product ij of the nodal block under the condition of considering neutron capture. U express 239 The decay heat contribution of U, P Np express 239 The decay heat contribution of Np.

[0192] This embodiment also provides a formula for calculating the nuclear power of a node, as shown in the following formula (16):

[0193]

[0194] Among them, P ijk Indicates the core power of the node. This represents the fission power of the segment. This represents the decay heat of the nodule.

[0195] In step S702 of some embodiments, the decay heat of the first target node is obtained based on the calculation formula of the decay heat of the node, and the sum of the decay heat and fission power of the first target node is calculated to obtain the nuclear power of the first target node.

[0196] In some embodiments, the core steady-state simulation results may also include the core's total fission power, total decay heat, or total nuclear power.

[0197] The total nuclear power is obtained by summing the nuclear power of each node in the reactor core. The total fission power P f It can be calculated using the following formula (17):

[0198]

[0199] The total decay heat P is obtained by summing the decay heat of each node in the reactor core. dIt can be calculated using the following formula (18):

[0200]

[0201] The total nuclear power 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 (19):

[0202]

[0203] Please see Figure 8 In some embodiments, the core steady-state simulation results include input values ​​from at least one partition of an external neutron detector, which can be the result of core neutron leakage to the external neutron detector.

[0204] The above step S102 may include, but is not limited to, steps S801 to S802:

[0205] Step S801: For at least one partition of the external neutron detector, the first average neutron flux density of each neutron energy group of the measurement block is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core.

[0206] Step S802: For each partition of at least one partition, determine the sixth target value of the target measurement block based on the average neutron flux density of the hot group of the target measurement block corresponding to the partition, the sensitivity coefficient of the off-pile neutron detector to the target measurement block, and the length of the partition.

[0207] Step S803: For each partition of at least one partition, determine the input value of the partition based on the sixth target value of each measurement block corresponding to the partition.

[0208] The target measurement block mentioned above is any block among the measurement blocks corresponding to the partition.

[0209] This embodiment also provides a formula for calculating the input value of the partition of the off-pile neutron detector, as shown in the following formula (20):

[0210]

[0211] In the formula, A i，dect Represents the input value of the i-th partition; N represents the number of measurable nodes in the i-th partition; s n This represents the sensitivity coefficient of the neutron detector to the measurement node n; The heat group average neutron flux density of the measurement block n corresponding to the i-th partition; l i This represents the length of the i-th partition of the neutron detector. Here, i ranges from 1 to I, where I is the number of partitions in the off-pile neutron detector.

[0212] In this embodiment, This is denoted as the sixth target value.

[0213] The aforementioned measurement segments are those that can be measured by the off-pile neutron detector.

[0214] In step S801 of some embodiments, the first average neutron flux density of each neutron energy group of the measurement block is calculated by the method of steps S401-S403 described above.

[0215] In step S802 of some embodiments, for the i-th partition in at least one partition, for each measurement node measured in the i-th partition, the product of the heat cluster average neutron flux density of the measurement node, the sensitivity coefficient of the off-pile neutron detector to the measurement node, and the length of the i-th partition is calculated to determine the sixth target value of the measurement node.

[0216] In step S803 of some embodiments, the sum of the sixth target values ​​of each measurement block corresponding to the i-th partition is calculated to obtain the input value A of the i-th partition. i，dect .

[0217] For example, a neutron detector may have at least one partition, including partition A and partition B. Partition A may have measurable measurement segments, including segments A and B, and partition B may have measurable measurement segments, including segment C. For the input value of partition A, the product of the average neutron flux density of segment A, the sensitivity coefficient of the neutron detector to segment A, and the length of partition A is calculated to obtain the first target product of segment A. The product of the average neutron flux density of segment B, the sensitivity coefficient of the neutron detector to segment B, and the length of partition A is calculated to obtain the second target product of segment B. The sum of the first target product and the second target product is calculated to obtain the first input value of partition A. For the input value of partition B, the product of the average neutron flux density of segment C, the sensitivity coefficient of the neutron detector to segment C, and the length of partition B is calculated to obtain the third target product of segment C. The third target product is the second input value of partition B.

[0218] In this embodiment of the application, the average neutron flux density can be obtained more accurately through the neutron integral balance equation. Subsequently, the steady-state simulation results of each core can be calculated using the average neutron flux density, thereby obtaining more accurate core steady-state simulation results and improving the effectiveness of core steady-state simulation.

[0219] 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.

[0220] For ease of understanding, a specific embodiment will be used as an example:

[0221] Step 1: The core steady-state model receives the reset command from the simulation platform, performs initial loading of the core steady-state model, and establishes a communication interface with the simulation platform.

[0222] Step 2: Read the model input card to complete the initialization of the core steady-state model and establish the core simulation model of the nuclear reactor.

[0223] Step 3: The core steady-state 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, the steady-state calculation for the current moment is performed.

[0224] Step 4: The core steady-state 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 formulas (6), (7), and (8).

[0225] Step 5: Spatially discretize the neutron steady-state diffusion equation using the semi-analytical nodal method, perform volume integration within the nodal, and introduce a modified coupling factor to obtain the expression (1) of the neutron integral equilibrium equation for the entire core.

[0226] Given the estimated coupling correction factor, the average neutron flux density of the neutron energy group in the block is obtained by calculating Equation (1). Specifically, the average neutron source term in the block is calculated according to Equation (2), and the net neutron flux and its correction flux are solved by using the transverse integration method and the residual weight method, and then the new coupling correction factor is solved. Specifically, the correction coupling factor of the neutron energy group in each coordinate direction is calculated according to Equations (3), (4) and (5).

[0227] Repeat the above process until the neutron flux density calculation results satisfy the iterative convergence.

[0228] The steady-state value of the delayed neutron precursor nuclear density is calculated according to the above formula (9).

[0229] Step 6: Based on the neutron flux density in Step 5, calculate the steady-state value of the fission product concentration of each block at the current moment according to the above formulas (10), (11), (12) and (13).

[0230] Step 7: Calculate the fission power, decay heat and nuclear power of each segment according to the above formulas (14), (15) and (16).

[0231] Calculate the total fission power, decay heat, and nuclear power of the reactor core according to the above formulas (17), (18), and (19).

[0232] Step 8: Calculate the input values ​​of each partition of the neutron detector according to the above formula (20) to simulate the result of core neutron leakage to the external neutron detector.

[0233] Step 9: Synchronously update and modify parameters such as thermal parameters and control commands online through the simulation platform.

[0234] 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.

[0235] 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.

[0236] Please see Figure 9 This application also provides a core steady-state simulation device for a nuclear reactor, which can implement the above-mentioned core steady-state simulation method for a nuclear reactor. The device 90 includes:

[0237] The thermal-hydraulic model 901 is used to send the target thermal parameters of the nuclear reactor to the simulation platform at the first moment.

[0238] Simulation Platform 902: Used to send steady-state calculation commands and target thermal parameters for the nuclear reactor to the core simulation model.

[0239] Core simulation model 903 is used to respond to steady-state calculation commands and determine the core steady-state 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. Among them, the neutron integral equilibrium equation is spatially discretized to determine the neutron steady-state diffusion equation through a nonlinear semi-analytical nodal method, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal.

[0240] The specific implementation of the core steady-state simulation device for this nuclear reactor is basically the same as the specific implementation of the core steady-state simulation method for the nuclear reactor described above, and will not be repeated here.

[0241] 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 core steady-state simulation method for a nuclear reactor. This electronic device can be any smart terminal, including tablet computers, in-vehicle computers, etc.

[0242] Please see Figure 10 , Figure 10 The hardware structure of an electronic device according to another embodiment is illustrated. The electronic device includes:

[0243] 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.

[0244] The memory 102 can be implemented as a read-only memory (ROM), a static storage device, a dynamic storage device, or a 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 using the XXX method of the embodiments of this application.

[0245] Input / output interface 103 is used to implement information input and output;

[0246] 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.).

[0247] Bus 105 transmits information between various components of the device (e.g., processor 101, memory 102, input / output interface 103, and communication interface 104);

[0248] The processor 101, memory 102, input / output interface 103 and communication interface 104 are connected to each other within the device via bus 105.

[0249] This application also provides a computer-readable storage medium storing a computer program that, when executed by a processor, implements the above-described core steady-state simulation method for a nuclear reactor.

[0250] 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.

[0251] The core steady-state simulation method, device, electronic equipment, and storage medium for nuclear reactors provided in this application send steady-state calculation commands for the nuclear reactor to the core simulation model through a simulation platform. In response to the steady-state calculation commands, the target thermal parameters of the nuclear reactor at the first moment are sent to the core simulation model through a thermal-hydraulic model. Based on the target thermal parameters and the neutron integral equilibrium equation of the entire core, the core simulation model determines the core steady-state 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 neutron steady-state diffusion equation, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal. Thus, this invention provides a core steady-state model that can be instructed to perform core steady-state simulation through steady-state calculation commands. The core steady-state model, responding to the steady-state calculation commands, automatically and intelligently performs core steady-state calculations based on the target thermal parameters and obtains the calculation results, avoiding manual operation and improving computational efficiency. Furthermore, since the nonlinear semi-analytical nodal method has high accuracy and high computational efficiency, spatial discretization of the neutron steady-state diffusion equation using the nonlinear semi-analytical nodal method can yield a more accurate and efficient neutron integral equilibrium equation, which is beneficial for obtaining more effective core steady-state simulation results.

[0252] 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.

[0253] 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.

[0254] 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.

[0255] 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.

[0256] 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.

[0257] 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.

[0258] 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.

[0259] 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.

[0260] 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.

[0261] 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.

[0262] 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 core steady-state simulation method for 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 steady-state calculation commands for the nuclear reactor and the target thermal parameters to the core simulation model. The core simulation model responds to the steady-state calculation command and, based on the target thermal parameters and the neutron integral equilibrium equation of the nuclear reactor core, determines the core steady-state simulation result of the nuclear reactor at the first moment; wherein, the neutron integral equilibrium equation is spatially discretized to determine the neutron steady-state diffusion equation through a nonlinear semi-analytical nodal method, and the average neutron source term within the nodal in the neutron integral equilibrium equation is determined by volume integration within the nodal; The 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, the second average neutron flux density of the neutron energy group of the second block, and the average neutron source term within the neutron energy group of the first block; the second block is the adjacent block of the first block in each coordinate direction.

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: 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 correction coupling 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, the second average neutron flux density, and the average neutron source term within the block.

3. The method according to claim 2, 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 correction coupling 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 correction coupling factor of the first block in the target coordinate direction is determined.

4. The method according to claim 2, characterized in that, The core steady-state simulation results include the first average neutron flux density of the first neutron energy group of the first target block, wherein the first target block is any block in the core of the nuclear reactor, and the first neutron energy group is any neutron energy group of the first target block; The determination of the core steady-state 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, determine the first cross-sectional data of the macroscopic neutron removal cross-section of the first neutron energy group; Obtain the corrected coupling factor of the first neutron energy group in each coordinate direction, the target parameter, and the average neutron source term within the block; The first cross-sectional data of the first neutron energy group, the corrected coupling factor of the first neutron energy group in each coordinate direction, 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.

5. The method according to claim 1, characterized in that, The core steady-state simulation results include the steady-state value of the slow-emission neutron precursor nuclear density of the first target node, wherein the first target node is any node in the core of the nuclear reactor; The determination of the core steady-state 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, the first average neutron flux density of the neutron energy group of the first target segment is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core. For each neutron energy group of the first target block, the second cross-sectional data of the macroscopic fission cross section of the neutron energy group is determined according to the target thermal parameters; 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. Based on the first target value of each of the neutron energy groups, determine the steady-state value of the slow-emission neutron precursor nuclear density of the first target block.

6. The method according to claim 1, characterized in that, The core steady-state simulation results include the steady-state concentration values ​​of fission products of the first target node, wherein the first target node is any node in the core of the nuclear reactor, 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 steady-state 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, further includes: For each neutron energy group of the first target segment, the first average neutron flux density of the neutron energy group of the first target segment is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core. For each neutron energy group of the first target block, the third cross-sectional data of the macroscopic absorption cross-section of the fission products of the neutron energy group is determined according to the target thermal parameters, and the second target value of the fission products of the neutron energy group is determined according to the third cross-sectional data of the fission products and the first average neutron flux density of the neutron energy group. Obtain the fourth coefficient corresponding to the fission product; wherein, if the fission product is fission product 135I or fission product 149Pm, the fourth coefficient is determined by the relevant parameters of the fission product, the relevant parameters including fission yield and decay constant; if the fission product is fission product 135Xe, the fourth coefficient is determined by the fission yield of fission product 135I, the relevant parameters of fission product 135Xe, and a third target value, the third target value being based on the neutron energy groups of the first target block in the first... The cross-sectional data of the microscopic absorption cross-section of the fission product 135Xe at the first time and the first average neutron flux density of each neutron energy group are used to determine the fourth coefficient. If the fission product is fission product 149Sm, the fourth coefficient is determined by the fission yield of fission product 149Sm and the fourth target value of fission product 149Sm. The fourth target value is determined based on the cross-sectional data of the microscopic absorption cross-section of the fission product 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. The steady-state concentration of the fission products of the first target node is determined based on the second target value of the fission products of each of the neutron energy groups and the fourth coefficient.

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

8. The method according to claim 1, characterized in that, The core steady-state simulation results include the input values ​​of at least one section of the external neutron detector. The determination of the core steady-state 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, further includes: For the measurement nodes measured by at least one partition of the off-pile neutron detector, the first average neutron flux density of each neutron energy group of the measurement node is determined based on the target thermal parameters and the neutron integral balance equation of the reactor core. For each of the at least one partition, a sixth target value for the target measurement node is determined based on the hot-group average neutron flux density of the target measurement node corresponding to the partition, the sensitivity coefficient of the off-pile neutron detector to the target measurement node, and the length of the partition; wherein, the target measurement node is any node among the measurement nodes corresponding to the partition; For each partition of the at least one partition, the input value of the partition is determined based on the sixth target value of each measurement block corresponding to the partition.

9. An electronic device, characterized in that, The electronic device includes a memory and a processor. The memory stores a computer program, and when the processor executes the computer program, it implements the core steady-state simulation method for a nuclear reactor as described in any one of claims 1 to 8.

10. A computer-readable storage medium storing a computer program, characterized in that, When the computer program is executed by the processor, it implements the core steady-state simulation method for the nuclear reactor as described in any one of claims 1 to 8.

Images