Reactor core steady state simulation method of nuclear reactor and related device

Abstract

The embodiment of the invention provides a reactor core steady state simulation method of a nuclear reactor and a related device, the method is applied to electronic equipment, the electronic equipment is provided with a simulation platform, and the simulation platform is provided with a thermal hydraulic model and a reactor core simulation model of the nuclear reactor. The thermal hydraulic model sends target thermal parameters of the nuclear reactor at the first moment to the simulation platform, and the simulation platform sends a steady-state calculation instruction and the target thermal parameters to the reactor core simulation model; and the reactor core simulation model responds to the steady-state calculation instruction and determines a reactor core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and a neutron integral balance equation of the reactor core of the nuclear reactor. According to the embodiment of the invention, the reactor core steady-state model can be instructed to carry out reactor core steady-state simulation through the steady-state calculation instruction, the reactor core steady-state model intelligently realizes the functions of carrying out reactor core steady-state calculation according to the target thermal parameters and obtaining the calculation result, manual operation is avoided, and the calculation efficiency is improved.

Description

Technical Field

[0001] The present application relates to the field of artificial intelligence technology, and in particular to a nuclear reactor core steady-state simulation method and related devices. Background Art

[0002] At present, for the core steady-state calculation of a nuclear reactor, generally after obtaining the thermal parameters obtained by simulating the nuclear reactor on a simulation platform, the core steady-state analysis software function is manually operated to realize the core steady-state simulation based on the obtained thermal parameters, resulting in the core steady-state analysis efficiency being too slow and the core steady-state simulation data cannot be obtained in a timely manner. Summary of the invention

[0003] The main purpose of the embodiments of the present application is to propose a core steady-state simulation method and related devices for a nuclear reactor, aiming to improve the efficiency of the core steady-state simulation.

[0004] To achieve the above-mentioned purpose, a first aspect of an embodiment of the present application provides a nuclear reactor core steady-state simulation method, which is applied to an electronic device, wherein the electronic device is provided with a simulation platform, and the simulation platform is provided with a thermal hydraulic model and a core simulation model of the nuclear reactor, and the method comprises:

[0005] Sending target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal hydraulic model;

[0006] Sending a steady-state calculation instruction and the target thermal parameters for the nuclear reactor to the core simulation model through the simulation platform;

[0007] The core simulation model responds to the steady-state calculation instruction 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 balance equation of the core of the nuclear reactor; wherein the neutron integral balance equation spatially discretizes the neutron steady-state diffusion equation by a nonlinear semi-analytical block method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

[0008] In some embodiments, the above 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 node, the second average neutron flux density of the neutron energy group of the second node, and the intra-node average neutron source term of the neutron energy group of the first node; wherein the second node is an adjacent node of the first node in each coordinate direction;

[0009] Before the neutron integral balance equation based on the target thermal parameters and the core of the nuclear reactor, the method further includes:

[0010] For each of the second nodes, determining a first coefficient corresponding to a second average neutron flux density of the second node according to target parameters of the first node and the second node in the target coordinate direction corresponding to the second node and a modified coupling factor of the first node in the target coordinate direction; wherein the target parameters in the target coordinate direction include a first size and a diffusion coefficient of the first node in the target coordinate direction, and a second size of the second node in the target coordinate direction;

[0011] Determine the second coefficient corresponding to the first average neutron flux density of the first node according to the first coefficient corresponding to the second average neutron flux density of each of the second nodes and the neutron macroscopic removal cross section of the neutron energy group;

[0012] The neutron integral balance equation is determined according to each of the first coefficients, the second coefficients, the first average neutron flux density, the second average neutron flux density, and the intra-node average neutron source term.

[0013] In some embodiments, before determining the first coefficient corresponding to the second average neutron flux density of the second node according to the target parameters of the first node and the second node in the target coordinate direction corresponding to the second node and the corrected coupling factor of the first node in the target coordinate direction, the method further includes:

[0014] Determine the net neutron differential flow of the neutron energy group on the node surface of the first node in the target coordinate direction according to the diffusion coefficient of the first node in the target coordinate direction and the characteristic parameters of the first node and the second node corresponding to the target coordinate direction; wherein the characteristic parameters include the average neutron flux density and size;

[0015] Determine a net neutron correction flow of the neutron energy group on the node surface of the first node in the target coordinate direction according to the net neutron differential flow of the neutron energy group on the node surface of the first node in the target coordinate direction and the net neutron flow of the neutron energy group on the node surface in the target coordinate direction;

[0016] A correction coupling factor of the first node in the target coordinate direction is determined according to the net neutron correction flow and characteristic parameters of the first node and the second node corresponding to the target coordinate direction.

[0017] In some embodiments, the core steady-state simulation result includes a first average neutron flux density of a first neutron energy group of a 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 step of determining a core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and a neutron integral balance equation of the core of the nuclear reactor comprises:

[0019] Determining first cross-sectional data of a neutron macroscopic removal cross-sectional area of ​​the first neutron energy group according to the target thermal parameters;

[0020] Obtaining a modified coupling factor of the first neutron energy group in each coordinate direction, the target parameter, and an average neutron source term within the node;

[0021] The first cross-sectional data of the first neutron energy group, the modified coupling factors of the first neutron energy group in each coordinate direction, the target parameters and the average neutron source term in the node are input into the neutron integral balance equation to obtain the first average neutron flux density.

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

[0023] The step of determining a core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and a neutron integral balance equation of the core of the nuclear reactor comprises:

[0024] For each neutron energy group of the first target node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determining a first average neutron flux density of the neutron energy group of the first target node;

[0025] For each neutron energy group of the first target node, second cross-sectional data of the macroscopic fission cross section of the neutron energy group is determined according to the target thermal parameters; and a 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] According to the first target value of each of the neutron energy groups, a steady-state value of the delayed neutron precursor nuclear density of the first target node is determined.

[0027] In some embodiments, the core steady-state simulation result includes a steady-state value of the concentration of the fission product of the first target node, wherein the first target node is any node in the core of the nuclear reactor, and the fission product includes at least one of the fission product 135I, the fission product 135Xe, the fission product 149Pm and the fission product 149Sm;

[0028] The determining of the steady-state simulation result of the core of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor further includes:

[0029] For each neutron energy group of the first target node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determining a first average neutron flux density of the neutron energy group of the first target node;

[0030] For each neutron energy group of the first target node, third cross-sectional data of the macroscopic absorption cross-section of the fission products of the neutron energy group are determined according to the target thermal parameters, and second target values ​​of the fission products of the neutron energy group are 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] Obtaining a 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 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, relevant parameters of fission product 135Xe and a third target value, the third target value being determined in the first target node according to each neutron energy group of the first target node. The fourth coefficient is determined by the cross-sectional data of the microscopic absorption cross-section of the fission product 135Xe at the first moment and the first average neutron flux density of each neutron energy group; if the fission product is the fission product 149Sm, the fourth coefficient is determined by the fission yield of the fission product 149Sm and the fourth target value of the fission product 149Sm, and the fourth target value is determined according to 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 moment and the first average neutron flux density of each neutron energy group;

[0032] The steady-state concentration value of the fission product of the first target node is determined according to the second target value of the fission product of each neutron energy group and the fourth coefficient.

[0033] In some embodiments, the core steady-state simulation result includes the fission power of the first target node, wherein the first target node is any node in the core of the nuclear reactor;

[0034] The determining of the steady-state simulation result of the core of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor further includes:

[0035] For each neutron energy group of the first target node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determining a first average neutron flux density of the neutron energy group of the first target node;

[0036] For each neutron energy group of the first target node, fourth cross-sectional data of the macroscopic energy generation cross-sectional area of ​​the neutron energy group is determined according to the target thermal parameters; and a 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 node;

[0037] determining the fission power of the first target node according to the fifth target value of each neutron energy group of the first target node;

[0038] The core steady-state simulation result includes the nuclear power of the first target node, and after determining the fission power of the first target node according to the fifth target value of each neutron energy group of the first target node, the method further includes:

[0039] The decay heat of the first target node is acquired, and the nuclear power of the first target node is obtained according to the decay heat of the first target node and the fission power of the first target node.

[0040] In some embodiments, the core steady-state simulation result includes an input value of at least one partition of an external neutron detector, and the determining of the core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor further includes:

[0041] For the measurement node measured by the at least one partition of the ex-core neutron detector, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determine the first average neutron flux density of each neutron energy group of the measurement node;

[0042] For each partition of the at least one partition, determine the sixth target value of the target measurement node according to the thermal group average neutron flux density of the target measurement node corresponding to the partition, the sensitivity coefficient of the ex-core 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, an input value of the partition is determined according to the sixth target value of each measurement node corresponding to the partition.

[0044] To achieve the above-mentioned purpose, a second aspect of an embodiment of the present application provides a nuclear reactor core steady-state simulation device, the device comprising:

[0045] The 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 for sending steady-state calculation instructions and target thermal parameters for the nuclear reactor to the core simulation model.

[0047] A core simulation model is used to respond to the steady-state calculation instruction and determine the core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor; wherein the neutron integral balance equation spatially discretizes the neutron steady-state diffusion equation by a nonlinear semi-analytical block method, and the average neutron source term in the block in the neutron integral balance equation is determined by volume integration in the block. .

[0048] To achieve the above objectives, a third aspect of an embodiment of the present application proposes an electronic device, which includes a memory and a processor, wherein the memory stores a computer program, and when the processor executes the computer program, the method described in the first aspect is implemented.

[0049] To achieve the above-mentioned purpose, the fourth aspect of an embodiment of the present application proposes a computer-readable storage medium, wherein the computer-readable storage medium stores a computer program, and when the computer program is executed by a processor, the method described in the first aspect is implemented.

[0050] The present application proposes a nuclear reactor core steady-state simulation method and related devices, the method is applied to electronic equipment, the electronic equipment is provided with a simulation platform, the simulation platform is built with 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 the steady-state calculation instructions and target thermal parameters for the nuclear reactor to the core simulation model, the core simulation model responds to the steady-state calculation instructions, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determines the steady-state simulation result of the core of the nuclear reactor at the first moment; wherein the neutron integral balance equation spatially discretizes the neutron steady-state diffusion equation by a nonlinear semi-analytical block method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block. Thus, the present invention provides a core steady-state model, which can realize the steady-state calculation instruction to instruct the core steady-state model to perform core steady-state simulation. The core steady-state model responds to the steady-state calculation instruction, automatically and intelligently realizes the function of performing core steady-state calculation according to target thermal parameters and obtaining calculation results, thereby avoiding manual operation and improving calculation efficiency. In addition, since the nonlinear semi-analytical block method has high accuracy and high calculation efficiency, the neutron steady-state diffusion equation is spatially discretized by the nonlinear semi-analytical block method, and a more accurate and efficient neutron integral balance equation can be obtained, which is conducive to obtaining more effective core steady-state simulation results later. BRIEF DESCRIPTION OF THE DRAWINGS

[0051] Figure 1 It is a flow chart of a nuclear reactor core steady-state simulation method provided in an embodiment of the present application;

[0052] Figure 2 yes Figure 1 Flow chart of the steps before step S103 in;

[0053] Figure 3 yes Figure 2 Flowchart of the steps before step S201 in;

[0054] Figure 4 yes Figure 1 Flow chart of step S103 in FIG.

[0055] Figure 5 yes Figure 1 Flow chart of step S103 in FIG.

[0056] Figure 6 yes Figure 1 Flow chart of step S103 in FIG.

[0057] Figure 7 yes Figure 1 Flow chart of step S103 in FIG.

[0058] Figure 8 yes Figure 1 Flow chart of step S103 in FIG.

[0059] Fig. 9 It is a structural schematic diagram of a nuclear reactor core steady-state simulation device provided in an embodiment of the present application;

[0060] Fig.10 It is a schematic diagram of the hardware structure of the electronic device provided in the embodiment of the present application. DETAILED DESCRIPTION

[0061] In order to make the purpose, technical solution and advantages of the present application more clearly understood, the present application is further described in detail below in conjunction with the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are only used to explain the present application and are not used to limit the present application.

[0062] It should be noted that, although the functional modules are divided in the device schematic diagram and the 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 above 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 those commonly understood by those skilled in the art to which this application belongs. The terms used herein are only for the purpose of describing the embodiments of this application and are not intended to limit this application.

[0064] The embodiments of the present application provide a core steady-state simulation method and device, electronic equipment and storage medium for a nuclear reactor, aiming to improve the efficiency of core steady-state simulation.

[0065] The steady-state simulation method and device, electronic device and storage medium of the core of a nuclear reactor provided in the embodiments of the present application are specifically explained through the following embodiments. First, the steady-state simulation method of the core of a nuclear reactor in the embodiments of the present application is described.

[0066] The core steady-state simulation method of a nuclear reactor provided in an embodiment of the present application relates to the field of artificial intelligence technology. The core steady-state simulation method of a nuclear reactor provided in an embodiment of the present application can be applied to a terminal, can also be applied to a server, and can also be software running in a terminal or a server. In some embodiments, the terminal can be an electronic device such as a smart phone, a tablet computer, a laptop computer, a desktop computer, etc.

[0067] Figure 1It is an optional flow chart of a steady-state simulation method for a core of a nuclear reactor provided in an embodiment of the present application. The method is applied to an electronic device, which is provided with a simulation platform, and the simulation platform is used to simulate a nuclear reactor. In some embodiments, the simulation platform is built with a thermal-hydraulic model and a core simulation model of the nuclear reactor. The thermal-hydraulic model is used to obtain thermal parameters of the nuclear reactor. The thermal parameters may be coolant temperature, effective fuel temperature, coolant density or soluble poison density, etc. The core simulation model is used to perform a steady-state simulation of a nuclear reactor.

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

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

[0070] Step S102, sending steady-state calculation instructions 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 instruction through the core simulation model, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determines the steady-state simulation result of the core of the nuclear reactor at the first moment.

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

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

[0074] In step S102 of some embodiments, when the core simulation model is required to run a core steady-state simulation of a nuclear reactor at a first moment according to demand, the operator can select an operation instruction in the operation interface to control the simulation platform, and the operation instruction is used to instruct the simulation platform to send steady-state calculation instructions 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 instruction, the simulation platform sends steady-state calculation instructions for the nuclear reactor and the target thermal parameters of the nuclear reactor at the first moment to the core simulation model.

[0075] Among them, the above-mentioned first moment can be the latest moment (current moment), which can realize real-time instructing 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.

[0076] In some embodiments, before step S103, a core steady-state model is first established, the core steady-state model receives a reset instruction from the simulation platform, and based on the reset instruction, the core steady-state model is reset (initial loading), a communication interface with the simulation platform is established, and a model input card is read, and the core steady-state model is initialized based on the model input card, and a core simulation model of the nuclear reactor is established, wherein the above-mentioned model input card may include core layout, node size or node division, etc.

[0077] The communication interface is used to receive control instructions and thermal parameters sent by the simulation platform in real time. The control instructions are used to instruct the core steady-state model to perform corresponding operations, wherein the control instructions include freeze instructions, run instructions (steady-state calculation instructions), storage instructions or backtracking instructions, etc. The freeze instructions are used to instruct the core steady-state model not to perform any processing, the storage instructions are used to instruct the core steady-state model to output the calculation results of the core steady-state model to the IC file, and the backtracking instructions are used to instruct the core steady-state model to perform the operation backtracking function. For example, the simulation platform sends the steady-state calculation instructions 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 spatially discretizes the neutron steady-state diffusion equation by a nonlinear semi-analytical block method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

[0079] The core simulation model receives the steady-state calculation instruction sent from the simulation platform through the communication interface, responds to the steady-state calculation instruction, and calculates the steady-state simulation result of the core of the nuclear reactor at the first moment according to the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor.

[0080] Steps S101 to S103 shown in the embodiment of the present application, send steady-state calculation instructions for the nuclear reactor to the core simulation model through the simulation platform, so that the core simulation model can respond to the steady-state calculation instructions to intelligently perform the core steady-state simulation function, without the need for manual core steady-state simulation operations, thereby improving the efficiency of the core steady-state simulation. Secondly, since the nonlinear semi-analytical block method has high accuracy and high computational efficiency, the neutron steady-state diffusion equation is spatially discretized by the nonlinear semi-analytical block method, and a more accurate and efficient neutron integral balance equation can be obtained, which is conducive to obtaining more effective core steady-state simulation results later. In addition, the simulation platform can send steady-state calculation instructions in real time, instructing the core simulation model to perform core steady-state simulation based on the latest thermal parameters, and the calculation speed can meet the requirements of real-time performance, and the process of the core steady-state simulation method can meet the range of various operating conditions, and has a wider application value.

[0081] In some embodiments, the above 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 node, the second average neutron flux density of the neutron energy group of the second node, and the average neutron source term within the node of the neutron energy group of the first node, wherein the second node is the adjacent node of the first node in each coordinate direction, see Figure 2 In some embodiments, a neutron integral balance 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 node, determine a first coefficient corresponding to a second average neutron flux density of the second node according to target parameters of the first node and the second node in the target coordinate direction corresponding to the second node, and a modified coupling factor of the first node in the target coordinate direction.

[0083] Step S202, determining the second coefficient corresponding to the first average neutron flux density of the first node according to the first coefficient corresponding to the second average neutron flux density of each second node and the neutron macroscopic removal cross section of the neutron energy group.

[0084] Step S203, determining a neutron integral balance equation according to each 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 node.

[0085] The above-mentioned coordinate directions include the positive and negative directions of the x-axis, the positive and negative directions of the y-axis or the positive and negative directions of the z-axis in the Cartesian coordinate system. The second node block may include the previous node block and the next node block of the first node block on the x-axis, the previous node block and the next node block of the first node block on the y-axis, and the previous node block and the next node block of the first node block on the z-axis.

[0086] In this embodiment, the above 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 include the first size and diffusion coefficient of the first node in the target coordinate direction. The target coordinate direction is the coordinate direction corresponding to the second node. For example, if the second node is the previous node of the first node on the x-coordinate axis, the target coordinate direction is the positive direction of the x-coordinate axis.

[0088] In step S201 of some embodiments, the first coefficient corresponding to the second average neutron flux density of each second node in the second node is determined. Specifically, for the first coefficient of the second node corresponding to the target coordinate direction which is the positive direction, the expression of the first coefficient is as follows:

[0089]

[0090] Among them, Cf 1 represents the first coefficient, h represents the node size, D represents the diffusion coefficient, represents the modified coupling factor, g represents the neutron energy group, u represents the target coordinate axis corresponding to the target coordinate direction, u can be x, y or z, u+ represents the target coordinate direction, that is, the positive direction of the u coordinate axis, h u Indicates the first size of the first node on the target coordinate axis corresponding to the target coordinate direction u+. Indicates the diffusion coefficient of the neutron energy group in the target coordinate direction u+, h' u Indicates the second size of the second node on the target coordinate axis corresponding to the target coordinate direction u+. Represents the corrected coupling factor of the neutron energy group in the target coordinate direction u+.

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

[0092]

[0093] Among them, Cf 1 represents the first coefficient, h represents the node size, D represents the diffusion coefficient, represents the modified coupling factor, g represents the neutron energy group, u represents the target coordinate axis corresponding to the target coordinate direction, u can be x, y or z, u- represents the target coordinate direction, that is, the negative direction of the target coordinate axis, h u Indicates the first size of the first node on the target coordinate axis corresponding to the target coordinate direction u-. Indicates the diffusion coefficient of the neutron energy group in the target coordinate direction u-, h' u Indicates the second size of the second node on the target coordinate axis corresponding to the target coordinate direction u-. It 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 node can be determined based on the first coefficient corresponding to the second average neutron flux density of each second node and the sum of the neutron macroscopic removal cross sections of the neutron energy group.

[0095] In one implementation, the expression of the second coefficient may be as follows:

[0096]

[0097] in, Cf2represents the second coefficient, ijk indicates the first section block, i+1jk indicates the previous section block of the first section block on the x-axis, i-1jk indicates the next section block of the first section block on the x-axis, ij+1k indicates the previous section block of the first section block on the y-axis, ij-1k indicates the next section block of the first section block on the y-axis, ijk+1 indicates the previous section block of the first section block on the z-axis, ijk-1 indicates the next section block of the first section block on the z-axis, Represents the neutron macroscopic removal cross section of the neutron energy group of the first block.

[0098] In step S203 of some embodiments, a neutron integral balance equation is constructed with the sum of the first product and each second product being equal to the average neutron source term in the nodal block of the neutron energy group, wherein 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 of the above neutron integral balance equation is shown in the following formula (1):

[0100]

[0101] in, represents the first average neutron flux density, represents the second average neutron flux density of the neutron energy group of the previous block on the x-axis of the first block, represents the second average neutron flux density of the neutron energy group of the next block after the first block on the x-axis, represents the second average neutron flux density of the neutron energy group of the previous block on the y-axis of the first block, It represents the second average neutron flux density of the neutron energy group of the next block after the first block on the y-coordinate axis, represents the second average neutron flux density of the neutron energy group of the previous block of the first block on the z coordinate axis, represents the second average neutron flux density of the neutron energy group of the next block after the first block on the z-coordinate axis, Represents the intra-block average neutron source term for the neutron energy group of the first block.

[0102] The calculation formula of the average neutron source term in the above-mentioned node block can be shown as the following formula (2):

[0103]

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

[0105] See also Figure 3In some embodiments, the steps before step S201 may include but are not limited to steps S301 to S304:

[0106] Step S301, determining the net neutron differential flow of the neutron energy group on the node surface of the first node in the target coordinate direction according to the diffusion coefficient of the first node in the target coordinate direction and the characteristic parameters of the first node and the second node corresponding to the target coordinate direction.

[0107] Step S302, determining the net neutron correction flow of the neutron energy group on the node surface of the first node in the target coordinate direction according to the net neutron differential flow of the neutron energy group on the node surface of the first node in the target coordinate direction and the net neutron flow of the neutron energy group on the node surface in the target coordinate direction.

[0108] Step S303, determining a correction coupling factor of the first node in the target coordinate direction according to the net neutron correction flux and characteristic parameters of the first node and the second node corresponding to the target coordinate direction.

[0109] The above 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 the net neutron flux and its correction flux, and then solve the new coupling correction factor.

[0111] In step S301 of some embodiments, the expression of the net neutron differential flow of the neutron energy group on the node surface of the first node in the target coordinate direction can be shown as the following formula (3):

[0112]

[0113] in, represents the net neutron differential flux of the neutron energy group on the surface of the first node in the target coordinate direction, Represents the average neutron flux density of the neutron energy group of the second node corresponding to the target coordinate direction.

[0114] In step S302 of some embodiments, the expression of the net neutron correction flux of the neutron energy group on the node surface of the first node in the target coordinate direction is shown in the following formula (4):

[0115]

[0116] in, represents the net neutron correction flux of the neutron energy group on the surface of the first node in the target coordinate direction, represents the net neutron flux of the neutron energy group on the surface of the first node in the target coordinate direction, It represents the net neutron difference flux of the neutron energy group on the block surface of the first block in the target coordinate direction.

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

[0118]

[0119] in, Represents the modified coupling factor of the first node in the target coordinate direction.

[0120] See also Figure 4 In some embodiments, the core steady-state simulation result includes a first average neutron flux density of a first neutron energy group of a first target node, the first target node is any node in the core of a 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, determining first cross-sectional data of a neutron macroscopic removal cross-sectional area of ​​a first neutron energy group according to target thermal parameters.

[0122] Step S402, obtaining the corrected coupling factor, target parameters and average neutron source term within the node of the first neutron energy group in each coordinate direction.

[0123] Step S403, inputting the first cross-sectional data of the first neutron energy group, the modified coupling factors of the first neutron energy group in each coordinate direction, the target parameters and the average neutron source term in the node into the neutron integral balance equation to obtain the first average neutron flux density.

[0124] In step S401 of some embodiments, this embodiment provides a calculation formula for calculating the cross-sectional data of each cross-sectional type, wherein the cross-sectional type can be divided into a neutron macroscopic removal cross-sectional area, a macroscopic fission cross-sectional area, etc. First, the cross-sectional data before the control rod correction is calculated, and then the cross-sectional data before the control rod correction is updated according to the rod position of the control rod to obtain the final cross-sectional data of the neutron energy group.

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

[0126]

[0127] In the formula, ijk indicates the node; n indicates the power of the expansion term; X indicates the cross-section type; g indicates the neutron energy group; It represents the cross-sectional data before the control rod is modified; Indicates the Taylor expansion coefficient of the X-type section; ΔT c , Δρ C , ΔN sp They respectively represent the changes in coolant temperature, effective fuel temperature, coolant density, and soluble poison (boric acid) density relative to the reference operating conditions.

[0128] Then, the cross-sectional data at the first moment calculated by the above formula (6) are updated according to the rod positions of each group of control rods to obtain the final cross-sectional data of the neutron energy group, wherein the cross-sectional data calculated by the formula (6) is updated according to the rod positions of each group of control rods by the following formula (7):

[0129]

[0130] In the formula, It represents the cross-sectional data after the control rod is corrected; represents the cross section of the control rod before modification; It represents the cross-sectional change introduced by the control rod being fully inserted into the segment; Represents the weight factor of the control rod inserted into the node.

[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-section data of the macroscopic absorption cross section at the first moment is updated according to the calculation result of the fission products (135Xe, 149Sm) at the first moment by the following formula (8):

[0132]

[0133] In the formula, represents the macroscopic absorption cross section after correction for fission products; represents the macroscopic absorption cross section before correction for fission products; are the microscopic absorption cross sections of 135Xe and 149Sm at the first moment; 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 of each node in the core of the nuclear reactor can be calculated by the above cross-sectional data calculation formulas (6) and (7).

[0135] In this embodiment, any node in the core of the nuclear reactor is recorded as the first target node, and any neutron energy group of the first target node is recorded as the first neutron energy group. The first target node and the first neutron energy group are used as objects to illustrate step S401-step S402. 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-sectional area of ​​the first neutron energy group of the first target node is determined according to 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 the aforementioned steps S301 to 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 the aforementioned steps S301 to S303 is an expression with only the first average neutron flux density of the first neutron energy group as an unknown quantity.

[0137] In step S403 of some embodiments, the first cross-sectional data of the first neutron energy group, the corrected coupling factors of the first neutron energy group in each coordinate direction, the target parameters in each coordinate direction and the average neutron source term in the node are input into the neutron integral balance equation, and a neutron integral balance equation with only the average neutron flux density as an unknown quantity can be obtained. The neutron integral balance equation is solved to obtain the first average neutron flux density of the first neutron energy group.

[0138] See also Figure 5 In some embodiments, the core steady-state simulation result includes a steady-state value of the delayed neutron precursor nuclear density of the first target node. 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 node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determine the first average neutron flux density of the neutron energy group of the first target node.

[0140] Step S502, for each neutron energy group of the first target node, 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, determining the steady-state value of the delayed neutron precursor nuclear density of the first target node according to the first target value of each neutron energy group.

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

[0143]

[0144] In the formula, C m represents the density of the mth group of delayed neutron precursors, G is the group number of the neutron energy group, represents the macroscopic fission cross section of the g′ group of the nodal block, represents the average neutron flux density of the g′ group of the node, β m represents the proportion of the mth group of delayed neutrons in all fission neutrons, λ m represents the decay constant of the mth group of delayed neutron precursor nuclei.

[0145] In this embodiment, the neutron energy group can be divided according to the thermal group and the fast group, so G can be taken as 2. Since the density of each group of delayed neutron pioneer nuclei is calculated based on the macroscopic fission cross section and the average neutron flux density of the neutron energy group of the same node, in this embodiment, the density of each group of delayed neutron pioneer nuclei is recorded as the delayed neutron pioneer nucleus density of the corresponding node.

[0146] In this embodiment, Recorded as the first target value.

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

[0148] In step S502 of some embodiments, second cross-sectional data of the macroscopic fission cross-section of each neutron energy group of the first target node are calculated according to the calculation formulas (6) and (7) for the above 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 mth group of delayed neutrons in all fission neutrons and the decay constant of the mth group of delayed neutron precursor nuclei corresponding to the first target node are obtained, and according to the above formula (9), the steady-state value of the delayed neutron precursor nucleus density of the first target node is calculated based on the proportion of the mth group of delayed neutrons in all fission neutrons and the decay constant of the mth group of delayed neutron precursor nuclei corresponding to the first target node and the first target values ​​of each neutron energy group corresponding to the first target node.

[0150] See also Figure 6In some embodiments, the core steady-state simulation result includes a steady-state value of the concentration of the fission product of the first target node, wherein the fission product includes at least one of the fission product 135I, the fission product 135Xe, the fission product 149Pm and the 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 node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determine the first average neutron flux density of the neutron energy group of the first target node.

[0152] Step S602, for each neutron energy group of the first target node, 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, obtaining the fourth coefficient corresponding to the fission product.

[0154] Step S604, determining the steady-state concentration value of the fission product of the first target node according to the second target value of the fission product of each neutron energy group and the fourth coefficient.

[0155] In this embodiment, a calculation formula for the steady-state value of the fission product concentration of each node is provided, wherein the calculation formula for the steady-state value of the fission product 135I concentration is shown in the following formula (10):

[0156]

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

[0158] The calculation formula for the steady-state value of the fission product 135Xe concentration is shown in the following formula (11):

[0159]

[0160] in, represents the steady-state value of the fission product 135Xe concentration in the nodule, γ I represents the fission yield of fission product 135I, γ Xerepresents the fission yield of the fission product 135Xe, λ Xe represents the decay constant of the fission product 135Xe, represents the macroscopic absorption cross section of the fission product 135Xe of the nodal g′ group, represents the average neutron flux density of the node g′ group, It represents the microscopic absorption cross section of the fission product 135Xe of the block g′ group at the first moment.

[0161] The calculation formula for the steady-state value of the fission product 149Pm concentration is shown in the following formula (12):

[0162]

[0163] in, represents the steady-state value of the fission product 149Pm concentration in the node, γ Pm represents the fission yield of the fission product 149Pm, λ Pm represents the decay constant of the fission product 149Pm, represents the macroscopic absorption cross section of the fission product 149Pm of the nodal g′ group, Represents the average neutron flux density of the node g′ group.

[0164] The calculation formula for the steady-state value of the fission product 149Sm concentration is shown in the following formula (13):

[0165]

[0166] in, represents the steady-state value of the fission product 149Sm concentration in the nodule, γ Pm represents the fission yield of the fission product 149Pm, represents the macroscopic absorption cross section of the fission product 149Sm of the nodal g′ group, represents the average neutron flux density of the node g′ group, It represents the microscopic absorption cross section of the fission product 149Sm of the block g′ group at the first moment.

[0167] In this embodiment, the calculation formulas for the steady-state values ​​of the concentrations of various fission products are divided by The part other than Recorded as the second target value.

[0168] Therefore, for the calculation formula of the steady-state value of the fission product 135I or the fission product 149Pm concentration, the fourth coefficient is determined by the relevant parameters of the fission product, which include the fission yield and the decay constant, that is, the fourth coefficient of the calculation formula of the steady-state value of the fission product 135I concentration is The fourth coefficient of the calculation formula for the steady-state concentration of the fission product 149Pm is

[0169] For the calculation formula of the steady-state value of the fission product 135Xe concentration, the fourth coefficient is determined by the fission yield of the fission product 135I, the relevant parameters of the fission product 135Xe, and the third target value. Determine, where the third target value According to the cross-sectional data of the microscopic absorption cross-section of the fission product 135Xe of 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 calculation formula of the steady-state value of the concentration of the fission product 135Xe is determined as follows:

[0170]

[0171] For the calculation formula of the steady-state value of the fission product 149Sm concentration, the fourth coefficient is obtained by dividing the fission yield of the fission product 149Sm by the fourth target value of the fission product 149Sm. OK, the fourth target value According to 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 moment and the first average neutron flux density of each neutron energy group, the fourth coefficient of the calculation formula of the steady-state value of the concentration of the fission product 149Sm is determined as follows:

[0172] In step S601 of some embodiments, first, the first average neutron flux density of each neutron energy group of the first target node is calculated according to the above steps S401 to 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 node is calculated according to the calculation formulas (6), (7), and (8) of the cross-sectional data. 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 product is obtained according to the corresponding fourth coefficient calculation method.

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

[0176] See also Figure 7In some embodiments, the core steady-state simulation result includes 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 node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determine the first average neutron flux density of the neutron energy group of the first target node.

[0178] Step S702, for each neutron energy group of the first target node, 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 node.

[0179] Step S703: determining the fission power of the first target node according to the fifth target value of each neutron energy group of the first target node.

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

[0181]

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

[0183] In this embodiment, Recorded as the fifth target value.

[0184] In step S701 of some embodiments, first, the first average neutron flux density of each neutron energy group of the first target node is calculated according to the above steps S401 to 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 node is calculated according to the calculation formulas (6) and (7) of the cross-sectional data. 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 node 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 the neutron energy groups of the first target node is calculated to obtain the fission power of the first target node.

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

[0188] Step S801, acquiring decay heat of a first target node, and obtaining the nuclear power of the first target node according to the decay heat of the first target node and the fission power of the first target node.

[0189] This embodiment provides a calculation formula for decay heat of a node, which is specifically shown in the following formula (15):

[0190]

[0191] Among them, P d (t) represents the decay heat of the node, ij represents the number of the fission product of the node, P dfij represents the decay heat of the fission product ij of the node under the consideration of neutron capture, P U express 239 Decay heat contribution of U, P Np express 239 Decay heat contribution of Np.

[0192] This embodiment also provides a calculation formula for the nuclear power of a node, which is specifically shown in the following formula (16):

[0193]

[0194] Among them, P ijk represents the nuclear power of the node, represents the fission power of the node, Represents the decay heat of the node.

[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 total fission power, total decay heat or total nuclear power of the core.

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

[0198]

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

[0200]

[0201] The total nuclear power is obtained by calculating the sum of the nuclear power of each node in the core. The total nuclear power P can be calculated by the following formula (19):

[0202]

[0203] See also Figure 8 In some embodiments, the core steady-state simulation result includes an input value of at least one partition of an ex-core neutron detector, and the input value of the ex-core neutron detector may be a result of core neutron leakage to the ex-core neutron detector.

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

[0205] Step S801, for a measurement node measured by at least one partition of an ex-core neutron detector, based on target thermal parameters and a neutron integral balance equation of a core of a nuclear reactor, determine a first average neutron flux density of each neutron energy group of the measurement node.

[0206] Step S802, for each partition of at least one partition, determine the sixth target value of the target measurement node according to the thermal group average neutron flux density of the target measurement node corresponding to the partition, the sensitivity coefficient of the ex-core neutron detector to the target measurement node, and the length of the partition.

[0207] Step S803: for each partition of at least one partition, determine the input value of the partition according to the sixth target value of each measurement node corresponding to the partition.

[0208] The target measurement node block is any node block in the measurement node blocks corresponding to the partition.

[0209] This embodiment also provides a calculation formula for the input value of the partition of the out-of-core neutron detector, which is specifically 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 nodes that can be measured in the i-th partition; s n It represents the sensitivity coefficient of the neutron detector to the measurement node n; represents the average neutron flux density of the thermal group corresponding to the measurement node n in the ith partition; l i Represents the length of the i-th partition of the neutron detector, where i ranges from 1 to I, and I is the number of partitions of the extra-pile neutron detector.

[0212] In this embodiment, Recorded as the sixth target value.

[0213] The above-mentioned measurement nodes are nodes that can be measured by the out-of-pile neutron detector.

[0214] In step S801 of some embodiments, the first average neutron flux density of each neutron energy group of the measurement node is calculated by the method of steps S401 to 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 thermal group average neutron flux density of the measurement node, the sensitivity coefficient of the ex-core 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 the measurement nodes corresponding to the i-th partition is calculated to obtain the input value A of the i-th partition. i，dect .

[0217] For example, at least one partition of the neutron detector includes partition A and partition B, and the measurement nodes that can be measured in partition A include nodes A and node B, and the measurement nodes that can be measured in partition B include node C. Then, for the input value of partition A, the product of the thermal group average neutron flux density of node A, the sensitivity coefficient of the neutron detector to node A, and the length of partition A is calculated to obtain the first target product of node A, and the product of the thermal group average neutron flux density of node B, the sensitivity coefficient of the neutron detector to node B, and the length of partition A is calculated to obtain the second target product of node 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 thermal group average neutron flux density of node C, the sensitivity coefficient of the neutron detector to node C, and the length of partition B is calculated to obtain the third target product of node C. The third target product is the second input value of partition B.

[0218] In an embodiment of the present application, a more accurate average neutron flux density can be obtained through the neutron integral balance equation, and then each core steady-state simulation result is calculated through the average neutron flux density, thereby obtaining a more accurate core steady-state simulation result, thereby improving the effectiveness of the core steady-state simulation.

[0219] It should be noted that the various optional implementation modes introduced in the embodiments of the present application can be implemented in combination with each other or separately if they do not conflict with each other, and the embodiments of the present application are not limited to this.

[0220] For ease of understanding, a specific embodiment is used as an example for illustration:

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

[0222] Step 2: The core steady-state model reads the model input card, completes the initialization of the core steady-state model, and establishes the core simulation model of the nuclear reactor.

[0223] Step 3: The core steady-state model receives control instructions from the simulation platform. If a freeze instruction is received, step 11 is directly executed; if a step or run instruction is received, the steady-state calculation at the current moment is performed.

[0224] Step 4: The core steady-state model receives the thermal parameters at the current moment fed back by 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: The neutron steady-state diffusion equation is spatially discretized by the semi-analytical nodal method, volume integration is performed within the nodal block, and a modified coupling factor is introduced to obtain the expression (1) of the neutron integral balance equation for the entire core.

[0226] Given the estimated coupling correction factor, the above formula (1) is calculated to obtain the average neutron flux density of the neutron energy group of the node. Among them, the average neutron source term in the node is calculated according to formula (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 corrected coupling factor of the neutron energy group in each coordinate direction is calculated according to formulas (3), (4) and (5).

[0227] Repeat the above process until the neutron flux density calculation results meet 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 node at the current moment according to the above equations (10), (11), (12) and (13).

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

[0231] According to the above formulas (17), (18) and (19), the total fission power, decay heat and nuclear power of the core are calculated respectively.

[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 thermal parameters, control instructions and other parameters online through the simulation platform.

[0234] Step 10: Determine whether a new control instruction from the simulation platform is received. If a storage instruction is received, the calculation result is output to the IC file; if a backtrace instruction is received, the backtrace function is executed.

[0235] Step 11: Repeat steps 3 to 10 according to the simulation platform control instructions until the operation end instruction is received.

[0236] See also Fig. 9 The embodiment of the present application further provides a nuclear reactor core steady-state simulation device, which can implement the above-mentioned nuclear reactor core steady-state simulation method. The device 90 includes:

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

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

[0239] The core simulation model 903 is used to respond to the steady-state calculation instruction and determine the steady-state simulation result of the core of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor; wherein the neutron integral balance equation spatially discretizes the neutron steady-state diffusion equation by a nonlinear semi-analytical block method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

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

[0241] The embodiment of the present application also provides an electronic device, the electronic device includes a memory and a processor, the memory stores a computer program, and the processor implements the above-mentioned nuclear reactor core steady-state simulation method when executing the computer program. The electronic device can be any smart terminal including a tablet computer, a car computer, etc.

[0242] See also Fig.10 , Fig.10 The hardware structure of an electronic device of another embodiment is illustrated, and the electronic device includes:

[0243] The processor 101 may be implemented by a general-purpose CPU (Central Processing Unit), a microprocessor, an application-specific integrated circuit (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 the present application;

[0244] The memory 102 may be implemented in the form of a read-only memory (ROM), a static storage device, a dynamic storage device, or a random access memory (RAM). The memory 102 may store an operating system and other application programs. When the technical solution provided in the embodiments of this specification is implemented by software or firmware, the relevant program code is stored in the memory 102, and the processor 101 calls and executes the XXX method of the embodiments of this application;

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

[0246] The communication interface 104 is used to realize the communication interaction between the device and other devices. The communication can be realized through a wired manner (such as USB, network cable, etc.) or a wireless manner (such as mobile network, WIFI, Bluetooth, etc.);

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

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

[0249] An embodiment of the present application also provides a computer-readable storage medium, which stores a computer program. When the computer program is executed by a processor, it implements the above-mentioned nuclear reactor core steady-state simulation method.

[0250] The memory, as a non-transient computer-readable storage medium, can be used to store non-transient software programs and non-transient computer executable programs. In addition, the memory may include a high-speed random access memory, and may also include a non-transient memory, such as at least one disk storage device, a flash memory device, or other non-transient solid-state storage device. In some embodiments, the memory may optionally include a memory remotely disposed relative to the processor, and these remote memories may be connected to the processor via a network. Examples of the above-mentioned network include, but are not limited to, the Internet, an intranet, a local area network, a mobile communication network, and combinations thereof.

[0251] The core steady-state simulation method, core steady-state simulation device, electronic device and storage medium of a nuclear reactor provided in the embodiment of the present application send a steady-state calculation instruction for a nuclear reactor to a core simulation model through a simulation platform, and send the target thermal parameters of the nuclear reactor at the first moment to the core simulation model through a thermal hydraulic model in response to the steady-state calculation instruction; the core steady-state simulation result of the nuclear reactor at the first moment is determined by the core simulation model based on the target thermal parameters and the neutron integral balance equation of the whole core; wherein the neutron integral balance equation discretizes the neutron steady-state diffusion equation in space by a nonlinear semi-analytical block method, and the average neutron source term in the block in the neutron integral balance equation is determined by volume integration in the block. In this way, the present invention provides a core steady-state model, which can realize the steady-state calculation instruction to instruct the core steady-state model to perform core steady-state simulation, and the core steady-state model responds to the steady-state calculation instruction, automatically and intelligently realizes the function of performing core steady-state calculation according to the target thermal parameters and obtaining the calculation result, avoiding manual operation and improving calculation efficiency. In addition, since the nonlinear semi-analytical nodal method has high accuracy and high computational efficiency, the neutron steady-state diffusion equation can be spatially discretized by the nonlinear semi-analytical nodal method to obtain a more accurate and efficient neutron integral balance equation, which is conducive to obtaining more effective core steady-state simulation results in the future.

[0252] The embodiments described in the embodiments of the present application are intended to more clearly illustrate the technical solutions of the embodiments of the present application and do not constitute a limitation on the technical solutions provided in the embodiments of the present application. Those skilled in the art will appreciate that with the evolution of technology and the emergence of new application scenarios, the technical solutions provided in the embodiments of the present application are also applicable to similar technical problems.

[0253] Those skilled in the art will appreciate that the technical solutions shown in the figures do not constitute a limitation on the embodiments of the present application, and may include more or fewer steps than shown in the figures, or a combination of certain steps, or different steps.

[0254] The device embodiments described above are merely illustrative, and the units described as separate components may or may not be physically separated, that is, they may be located in one place or distributed on multiple network units. Some or all of the modules may be selected according to actual needs to achieve the purpose of the solution of this embodiment.

[0255] Those skilled in the art will appreciate that all or some of the steps in the methods disclosed above, and the functional modules / units in the systems and devices may be implemented as software, firmware, hardware, or a suitable combination thereof.

[0256] The terms "first", "second", "third", "fourth", etc. (if any) in the specification of the present application and the above-mentioned drawings are used to distinguish similar objects, and are not necessarily used to describe a specific order or sequence. It should be understood that the data used in this way can be interchangeable where appropriate, so that the embodiments of the present application described herein can be implemented in an order other than those illustrated or described herein. In addition, the terms "including" and "having" and any of their variations are intended to cover non-exclusive inclusions, for example, a process, method, system, product or device comprising a series of steps or units is not necessarily limited to those steps or units clearly listed, but may include other steps or units that are not clearly listed or inherent to these processes, methods, products or devices.

[0257] It should be understood that in the present application, "at least one (item)" means one or more, and "plurality" means two or more. "And / or" is used to describe the association relationship of associated objects, indicating that three relationships may exist. For example, "A and / or B" can mean: only A exists, only B exists, and A and B exist at the same time, where A and B can be singular or plural. The character " / " generally indicates that the objects associated before and after are in an "or" relationship. "At least one of the following" or similar expressions refers to any combination of these items, including any combination of single or plural items. For example, at least one of a, b or c can mean: a, b, c, "a and b", "a and c", "b and c", or "a and b and c", where a, b, c can be single or multiple.

[0258] In the several embodiments provided in the present application, it should be understood that the disclosed devices and methods can be implemented in other ways. For example, the device embodiments described above are only schematic. For example, the division of the above units is only a logical function division. There may be other division methods in actual implementation, such as multiple units or components can be combined or integrated into another system, or some features can be ignored or not executed. Another point is that the mutual coupling or direct coupling or communication connection shown or discussed can be through some interfaces, indirect coupling or communication connection of devices or units, which can be electrical, mechanical or other forms.

[0259] The units described above as separate components may or may not be physically separated, and the components shown as units may or may not be physical units, that is, they may be located in one place or distributed on multiple network units. Some or all of the units may be selected according to actual needs to achieve the purpose of the solution of this embodiment.

[0260] In addition, each functional unit in each embodiment of the present application may be integrated into one processing unit, or each unit may exist physically separately, or two or more units may be integrated into one unit. The above-mentioned integrated unit may be implemented in the form of hardware or in the form of software functional units.

[0261] If the integrated unit is implemented in the form of 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 the present application, 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, and the computer software product is stored in a storage medium, including multiple instructions to enable a computer device (which can be a personal computer, server, or network device, etc.) to execute all or part of the steps of the methods of various embodiments of the present application. The aforementioned storage medium includes: U disk, mobile hard disk, read-only memory (Read-Only Memory, referred to as ROM), random access memory (Random Access Memory, referred to as RAM), disk or optical disk and other media that can store programs.

[0262] The preferred embodiments of the present invention are described above with reference to the accompanying drawings, but the scope of the rights of the present invention is not limited thereto. Any modification, equivalent substitution and improvement made by a person skilled in the art without departing from the scope and essence of the present invention should be within the scope of the rights of the present invention.

Claims

1. A nuclear reactor core steady-state simulation method, characterized in that: Applied to electronic equipment, the electronic equipment is provided with a simulation platform, and the simulation platform is built with a thermal hydraulic model and a core simulation model of the nuclear reactor; Sending target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal hydraulic model; Sending a steady-state calculation instruction and the target thermal parameters for the nuclear reactor to the core simulation model through the simulation platform; The core simulation model responds to the steady-state calculation instruction 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 balance equation of the core of the nuclear reactor; wherein the neutron integral balance equation spatially discretizes the neutron steady-state diffusion equation by a nonlinear semi-analytical block method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

2. The method according to claim 1, characterized in that 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 node, the second average neutron flux density of the neutron energy group of the second node, and the intra-node average neutron source term of the neutron energy group of the first node; wherein the second node is an adjacent node of the first node in each coordinate direction; Before the neutron integral balance equation based on the target thermal parameters and the core of the nuclear reactor, the method further includes: For each of the second nodes, determining a first coefficient corresponding to a second average neutron flux density of the second node according to target parameters of the first node and the second node in the target coordinate direction corresponding to the second node and a modified coupling factor of the first node in the target coordinate direction; wherein the target parameters in the target coordinate direction include a first size and a diffusion coefficient of the first node in the target coordinate direction, and a second size of the second node in the target coordinate direction; Determine the second coefficient corresponding to the first average neutron flux density of the first node according to the first coefficient corresponding to the second average neutron flux density of each of the second nodes and the neutron macroscopic removal cross section of the neutron energy group; The neutron integral balance equation is determined according to each of the first coefficients, the second coefficients, the first average neutron flux density, the second average neutron flux density, and the intra-node average neutron source term.

3. The method according to claim 2, characterized in that Before determining a first coefficient corresponding to a second average neutron flux density of the second node according to target parameters of the first node and the second node in a target coordinate direction corresponding to the second node and a modified coupling factor of the first node in the target coordinate direction, the method further includes: Determine the net neutron differential flow of the neutron energy group on the node surface of the first node in the target coordinate direction according to the diffusion coefficient of the first node in the target coordinate direction and the characteristic parameters of the first node and the second node corresponding to the target coordinate direction; wherein the characteristic parameters include the average neutron flux density and size; Determine a net neutron correction flow of the neutron energy group on the node surface of the first node in the target coordinate direction according to the net neutron differential flow of the neutron energy group on the node surface of the first node in the target coordinate direction and the net neutron flow of the neutron energy group on the node surface in the target coordinate direction; A correction coupling factor of the first node in the target coordinate direction is determined according to the net neutron correction flow and characteristic parameters of the first node and the second node corresponding to the target coordinate direction.

4. The method according to claim 2, characterized in that: The core steady-state simulation result includes a first average neutron flux density of a first neutron energy group of a 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; The step of determining a core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and a neutron integral balance equation of the core of the nuclear reactor comprises: Determining first cross-sectional data of a neutron macroscopic removal cross-sectional area of ​​the first neutron energy group according to the target thermal parameters; Obtaining a modified coupling factor of the first neutron energy group in each coordinate direction, the target parameter, and an average neutron source term within the node; The first cross-sectional data of the first neutron energy group, the modified coupling factors of the first neutron energy group in each coordinate direction, the target parameters and the average neutron source term in the node 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 result includes a 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; The step of determining a core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and a neutron integral balance equation of the core of the nuclear reactor comprises: For each neutron energy group of the first target node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determining a first average neutron flux density of the neutron energy group of the first target node; For each neutron energy group of the first target node, second cross-sectional data of the macroscopic fission cross section of the neutron energy group is determined according to the target thermal parameters; and a 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; According to the first target value of each of the neutron energy groups, a steady-state value of the delayed neutron precursor nuclear density of the first target node is determined.

6. The method according to claim 1, characterized in that The core steady-state simulation result includes a steady-state value of the concentration of the fission product of the first target node, wherein the first target node is any node in the core of the nuclear reactor, and the fission product includes at least one of the fission product 135I, the fission product 135Xe, the fission product 149Pm and the fission product 149Sm; The determining of the steady-state simulation result of the core of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor further includes: For each neutron energy group of the first target node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determining a first average neutron flux density of the neutron energy group of the first target node; For each neutron energy group of the first target node, third cross-sectional data of the macroscopic absorption cross-section of the fission products of the neutron energy group are determined according to the target thermal parameters, and second target values ​​of the fission products of the neutron energy group are determined according to the third cross-sectional data of the fission products and the first average neutron flux density of the neutron energy group; Obtaining a 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 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, relevant parameters of fission product 135Xe and a third target value, the third target value being determined in the first target node according to each neutron energy group of the first target node. The fourth coefficient is determined by the cross-sectional data of the microscopic absorption cross-section of the fission product 135Xe at the first moment and the first average neutron flux density of each neutron energy group; if the fission product is the fission product 149Sm, the fourth coefficient is determined by the fission yield of the fission product 149Sm and the fourth target value of the fission product 149Sm, and the fourth target value is determined according to 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 moment and the first average neutron flux density of each neutron energy group; The steady-state concentration value of the fission product of the first target node is determined according to the second target value of the fission product of each neutron energy group and the fourth coefficient.

7. The method according to claim 1, characterized in that The core steady-state simulation result includes the fission power of the first target node, wherein the first target node is any node in the core of the nuclear reactor; The determining of the steady-state simulation result of the core of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor further includes: For each neutron energy group of the first target node, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determining a first average neutron flux density of the neutron energy group of the first target node; For each neutron energy group of the first target node, fourth cross-sectional data of the macroscopic energy generation cross-sectional area of ​​the neutron energy group is determined according to the target thermal parameters; and a 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 node; determining the fission power of the first target node according to the fifth target value of each neutron energy group of the first target node; The core steady-state simulation result includes the nuclear power of the first target node, and after determining the fission power of the first target node according to the fifth target value of each neutron energy group of the first target node, the method further includes: The decay heat of the first target node is acquired, and the nuclear power of the first target node is obtained according to the decay heat of the first target node and the fission power of the first target node.

8. The method according to claim 1, characterized in that The core steady-state simulation result includes an input value of at least one partition of an external neutron detector, and the step of determining the core steady-state simulation result of the nuclear reactor at the first moment based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor further includes: For the measurement node measured by the at least one partition of the ex-core neutron detector, based on the target thermal parameters and the neutron integral balance equation of the core of the nuclear reactor, determine the first average neutron flux density of each neutron energy group of the measurement node; For each partition of the at least one partition, determine the sixth target value of the target measurement node according to the thermal group average neutron flux density of the target measurement node corresponding to the partition, the sensitivity coefficient of the ex-core 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, an input value of the partition is determined according to the sixth target value of each measurement node corresponding to the partition.

9. An electronic device, characterized in that: The electronic device comprises a memory and a processor, the memory stores a computer program, and the processor implements the nuclear reactor core steady-state simulation method according to any one of claims 1 to 8 when executing the computer program.

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

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