Reactor core transient simulation method of nuclear reactor, electronic equipment and medium

Abstract

The embodiment of the invention provides a reactor core transient simulation method of a nuclear reactor, electronic equipment and a medium, the method is applied to the 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. Target thermal parameters of the nuclear reactor at the first moment are sent to the simulation platform through the thermal hydraulic model; a transient calculation instruction and target thermal parameters for the nuclear reactor are sent to the reactor core simulation model through the simulation platform; and determining a reactor core transient simulation result of the nuclear reactor at the first moment based on the target thermal parameters and a neutron integral equilibrium equation of the reactor core of the nuclear reactor in response to the transient calculation instruction through the reactor core simulation model. According to the embodiment of the invention, the transient simulation of the reactor core can be performed by indicating the reactor core transient model through the transient calculation instruction, and the transient calculation of the reactor core is automatically performed by the reactor core transient model, so that the transient simulation efficiency of the reactor core of the nuclear reactor 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 transient simulation method, electronic equipment and medium. Background Art

[0002] During the design, operation and maintenance of nuclear reactors, transient calculation of the core of nuclear reactors is a key link in evaluating the safety and performance of nuclear reactors. Transient calculation of the core of nuclear reactors involves dynamic analysis of thermal-hydraulic parameters and neutron physical parameters of nuclear reactors under non-steady-state conditions.

[0003] However, the existing nuclear reactor core transient calculation generally simulates the nuclear reactor through a simulation platform. After obtaining the thermal-hydraulic parameters, the core transient analysis software function is manually operated to realize the nuclear reactor core transient simulation based on the obtained thermal-hydraulic parameters, resulting in too low efficiency of the nuclear reactor core transient analysis and failure to obtain the nuclear reactor core transient simulation data in a timely manner. Summary of the invention

[0004] The main purpose of the embodiments of the present application is to propose a nuclear reactor core transient simulation method, electronic equipment and medium, aiming to obtain the transient simulation data of the nuclear reactor core in a timely manner, thereby improving the transient simulation efficiency of the nuclear reactor core.

[0005] To achieve the above-mentioned purpose, a first aspect of an embodiment of the present application provides a nuclear reactor core transient 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:

[0006] Sending target thermal parameters of the nuclear reactor at the first moment to the simulation platform through the thermal hydraulic model;

[0007] Sending a transient calculation instruction and the target thermal parameters for the nuclear reactor to the core simulation model through the simulation platform;

[0008] The core simulation model responds to the transient calculation instruction and determines the transient 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 transient neutron diffusion equation by a nonlinear semi-analytical block method, the transient neutron diffusion equation is calculated by an implicit difference method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

[0009] To achieve the above-mentioned purpose, a second aspect of an embodiment of the present application provides a nuclear reactor core transient simulation device, 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 device comprises:

[0010] A first sending module, used for sending target thermal parameters of the nuclear reactor at a first moment to the simulation platform through the thermal hydraulic model;

[0011] A second sending module, used for sending a transient calculation instruction for the nuclear reactor and the target thermal parameters to the core simulation model through the simulation platform;

[0012] A transient simulation module is used to respond to the transient calculation instruction through the core simulation model, and determine the transient 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 transient neutron diffusion equation by a nonlinear semi-analytical block method, the transient neutron diffusion equation is calculated by an implicit difference method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

[0013] 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.

[0014] 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.

[0015] The present application proposes a nuclear reactor core transient simulation method, electronic equipment and medium. The method is applied to the electronic equipment. The electronic equipment is provided with a simulation platform. The simulation platform is equipped with a thermal-hydraulic model and a core simulation model of the nuclear reactor. The target thermal parameters of the nuclear reactor at the first moment are sent to the simulation platform through the thermal-hydraulic model; the transient calculation instructions and target thermal parameters for the nuclear reactor are sent to the core simulation model through the simulation platform; the core simulation model responds to the transient calculation instructions, and determines the transient 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; In the method, the neutron integral balance equation spatially discretizes the transient neutron diffusion equation by a nonlinear semi-analytical block method, the transient neutron diffusion equation is calculated by an implicit difference method, and the average neutron source term in the block in the neutron integral balance equation is determined by volume integration in the block; thus, the present invention provides a core transient model, which can realize the transient calculation instruction to instruct the core transient model to perform core transient simulation, and the core transient model responds to the transient calculation instruction, automatically and intelligently realizes the function of performing core transient calculation according to the target thermal parameters and obtaining the calculation result, thereby avoiding manual operation and improving calculation efficiency. In addition, the transient neutron diffusion equation is obtained by the implicit difference method, which can better deal with the temporal changes of the main parameters in the reactor core, reduce numerical instability and time step limitations, and improve the transient simulation accuracy of the reactor core; the transient neutron diffusion equation is then spatially discretized by the nonlinear semi-analytical block method, which can more accurately simulate the neutron behavior in the nuclear reactor core. Therefore, the transient neutron diffusion equation is obtained by the implicit difference method, and then the transient neutron diffusion equation is spatially discretized by the nonlinear semi-analytical method to obtain a neutron integral balance equation with more accurate and efficient calculation results, thereby further improving the accuracy of the core transient simulation results and improving the transient simulation efficiency of the reactor core. BRIEF DESCRIPTION OF THE DRAWINGS

[0016] Figure 1 It is a flow chart of a nuclear reactor core transient simulation method provided in an embodiment of the present application;

[0017] Figure 2 It is a structural schematic diagram of a nuclear reactor core transient simulation device provided in an embodiment of the present application;

[0018] Figure 3 It is a schematic diagram of the hardware structure of the electronic device provided in the embodiment of the present application. DETAILED DESCRIPTION

[0019] 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.

[0020] 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.

[0021] 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.

[0022] The embodiments of the present application provide a method, device, electronic device and storage medium for transient simulation of the core of a nuclear reactor, aiming to obtain transient simulation data of the core of a nuclear reactor in a timely manner, thereby improving the transient simulation efficiency of the core of a nuclear reactor.

[0023] The core transient simulation method, device, electronic device and storage medium of a nuclear reactor provided in the embodiments of the present application are specifically described through the following embodiments. First, the core transient simulation method of a nuclear reactor in the embodiments of the present application is described.

[0024] The core transient simulation method of a nuclear reactor provided in an embodiment of the present application relates to the field of artificial intelligence technology. The core transient 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.

[0025] Figure 1 It is an optional flow chart of a method for transient simulation of 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 constructed 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 transient simulation of the nuclear reactor.

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

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

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

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

[0030] In this step, when the core simulation model is required to run the core transient simulation of the nuclear reactor at the 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 transient 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 transient calculation instructions for the nuclear reactor and the target thermal parameters of the nuclear reactor at the first moment to the core simulation model.

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

[0032] Step S103, responding to the transient calculation 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, to determine the transient simulation result of the core of the nuclear reactor at the first moment; wherein the neutron integral balance equation spatially discretizes the transient neutron diffusion equation by a nonlinear semi-analytical block method, the transient neutron diffusion equation is calculated by an implicit difference method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

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

[0034] 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 transient model to run corresponding operations, wherein the control instructions include freeze instructions, run instructions (transient calculation instructions), storage instructions or backtracking instructions, etc. The freeze instructions are used to instruct the core transient model not to perform any processing, the storage instructions are used to instruct the core transient model to output the calculation results of the core transient model to the IC file, and the backtracking instructions are used to instruct the core transient model to perform the operation backtracking function. For example, the simulation platform sends the transient 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.

[0035] The above neutron integral balance equation spatially discretizes the transient neutron diffusion equation by a nonlinear semi-analytical nodal method, wherein the transient neutron diffusion equation temporally discretizes the space-time neutron dynamics equation group by an implicit difference method, and the nodal average neutron source term in the neutron integral balance equation is determined by volume integration within the nodal block.

[0036] The core simulation model receives transient calculation instructions sent from the simulation platform through the communication interface, responds to the transient calculation instructions, and calculates the transient simulation results 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.

[0037] Steps S101 to S103 shown in the embodiment of the present application send transient 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 transient calculation instructions to intelligently perform the core transient simulation function, without the need for manual core transient simulation operations, thereby improving the efficiency of the core transient simulation. Secondly, the transient neutron diffusion equation is obtained by the implicit difference method, which can better handle the time changes of the main parameters in the reactor core, reduce numerical instability and time step restrictions, so as to improve the transient simulation accuracy of the reactor core; and then the transient neutron diffusion equation is spatially discretized by the nonlinear semi-analytical block method, which can be more accurate. Accurately simulate the neutron behavior in the nuclear reactor core; therefore, the transient neutron diffusion equation is obtained by the implicit difference method, and then the transient neutron diffusion equation is spatially discretized by the nonlinear semi-analytical method to obtain a neutron integral balance equation with more accurate calculation results and higher efficiency, thereby further improving the accuracy of the core transient simulation results and improving the transient simulation efficiency of the reactor core; in addition, the simulation platform can send transient calculation instructions in real time, instructing the core simulation model to perform core transient simulation based on the thermal parameters at the latest moment, the calculation speed can meet the real-time requirements, and the process of the core transient simulation method can meet various operating conditions, and has a wider application value.

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

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

[0040] Step S202: Use the ratio of the first parameter to the neutron flux density as a first dynamic frequency representation.

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

[0042] Step S204: using the ratio of the second parameter to the neutron precursor nuclear density as a second dynamic frequency representation.

[0043] Step S205 : discretely process the first dynamic frequency and the second dynamic frequency respectively by implicit difference method to obtain a differential representation of the first dynamic frequency and a differential representation of the second dynamic frequency.

[0044] Step S206: Obtain the transient neutron diffusion equation according to the differential representation of the first dynamic frequency and the differential representation of the second dynamic frequency.

[0045] Step S207, spatially discretizing the transient neutron diffusion equation to obtain the neutron integral balance equation.

[0046] In this implementation, the core of the reactor is divided into multiple blocks according to the Cartesian coordinate system, and the dynamic frequency expressions of the neutron flux density in each block and the density of each group of delayed neutron precursor nuclei in the block that change with time are defined. The time terms in the dynamic frequency expression of the neutron flux density (i.e., the first dynamic frequency expression) and the dynamic frequency expression of the delayed neutron precursor nuclei (i.e., the second dynamic frequency expression) are discretized by the implicit difference method to obtain the differential expression of the dynamic frequency of the neutron flux density and the differential expression of the dynamic frequency of the delayed neutron precursor nuclei. The differential expression of the dynamic frequency of the neutron flux density and the differential expression of the dynamic frequency of the delayed neutron precursor nuclei are substituted into the space-time neutron dynamic equation group, the delayed precursor neutron nucleus concentration is eliminated, and the transient neutron diffusion equation is obtained. Then, the neutron integral balance equation is obtained by spatially discretizing the transient neutron diffusion equation.

[0047] Specifically, the dynamic frequency expression of the neutron flux density and the dynamic frequency expression of the delayed neutron precursor density can be expressed as follows:

[0048]

[0049] Among them, ω g,ijk represents the dynamic frequency of the neutron flux density of the g group in the ijk block, φ g,ijk represents the neutron flux density of group g in the ijk block, t represents the running time of the transient model, represents the dynamic frequency of the density of the mth group of delayed neutron precursor nuclei in the ijk block, ijk represents the block, g represents the neutron energy group, and m represents the number of delayed neutron precursor nuclei. represents the density of the mth group of delayed neutron precursor nuclei.

[0050] In this embodiment, Recorded as the first parameter, Denoted as the first dynamic frequency representation, Recorded as the second parameter, It is recorded as the second dynamic frequency representation.

[0051] By using implicit difference method to discretize the time term in the dynamic frequency expression of neutron flux density and the dynamic frequency expression of delayed neutron precursor nuclear density, the differential expression of the dynamic frequency of neutron flux density and the differential expression of the dynamic frequency of delayed neutron precursor nuclear density can be obtained. The differential expression of the dynamic frequency of neutron flux density and the differential expression of the dynamic frequency of delayed neutron precursor nuclear density can be expressed by the following formula (2):

[0052]

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

[0054] Substituting the differential expression of the dynamic frequency of the neutron flux density and the differential expression of the dynamic frequency of the delayed neutron precursor nuclear density into the space-time neutron dynamics equations, eliminating the delayed precursor neutron nuclear concentration, the transient neutron diffusion equation is obtained. The transient neutron diffusion equation can be expressed as follows (3):

[0055]

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

[0057] Furthermore, the calculation formula of the above neutron source term can be expressed as follows:

[0058]

[0059] Where G represents the group number of the neutron energy group, represents the macroscopic scattering cross section of neutrons from the g′ energy group to the g energy group, χ g,ijk represents the fraction of prompt neutrons of neutron energy group g in the ijk node produced by fission, represents the macroscopic fission cross section of the g′ neutron energy group in the ijk block, φ g′,ijk represents the neutron flux density of the g′ neutron energy group in the ijk node, and g′ represents the g′ neutron energy group.

[0060] The calculation expressions of the neutron macroscopic removal cross section of the g neutron energy group in the ijk node and the fraction of prompt neutrons produced by fission to the g neutron energy group in the ijk node are redefined. The above expressions can be expressed by the following formula (5):

[0061]

[0062] in, represents the neutron macroscopic fission cross section of the g neutron energy group in the redefined ijk block, represents the neutron macroscopic fission cross section of the g neutron energy group in the ijk block, v g represents the neutron velocity, and β represents the delayed neutron fraction of each group β m The sum of 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, represents the fraction of delayed neutrons produced by the mth group of precursor nuclei to the g energy group, Represents the fraction of prompt neutrons produced by the redefined fission into the g neutron energy group in the ijk node.

[0063] By redefining the neutron macroscopic removal cross section of the g neutron energy group in the ijk block and the calculation expression of the prompt neutron fraction produced by fission to the g neutron energy group in the ijk block, the transient neutron diffusion equation of formula (3) can be arranged into the same expression form as the steady-state neutron diffusion equation, so as to facilitate the subsequent spatial discretization of the transient neutron diffusion equation using the semi-analytical block method to obtain the neutron integral balance equation.

[0064] It should be noted that 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.

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

[0066] In this embodiment, by defining the dynamic frequency expressions of the neutron flux density and the neutron precursor nucleus, and then differentiating the time terms in the above dynamic frequency expressions through the implicit difference method, the differential expressions of the neutron flux density and the neutron precursor nucleus are obtained, and the above differential expressions are substituted into the space-time neutron dynamics equation group to obtain the transient neutron diffusion equation, which can better handle the temporal changes of the main parameters in the reactor core, reduce numerical instability and time step limitations, so as to improve the transient simulation accuracy of the reactor core.

[0067] In some embodiments, the above-mentioned neutron integral balance equation is used to characterize the relationship between the first average neutron flux density of the neutron energy group of the first node at the first moment, the second average neutron flux density of the neutron energy group of the second node at the first moment, 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. In some embodiments, the neutron integral balance equation is established before step S103, and the steps before step S103 may include but are not limited to steps S301 to S303.

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

[0069] Step S302: 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 second node and the neutron macroscopic removal cross section of the neutron energy group.

[0070] Step S303 determines the neutron integral balance equation according to each of the first coefficients, the second coefficients, the first average neutron flux density at the first moment, the second average neutron flux density at the first moment, and the average neutron source term within the node.

[0071] In this implementation, the above neutron integral balance equation is used to calculate the first average neutron flux density of the neutron energy group of the first node at the first moment.

[0072] 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.

[0073] 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:

[0074]

[0075] Wherein, Cf1 represents the first coefficient, h represents the node size, and D represents the diffusion coefficient. represents the coupling correction 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 coupling correction factor of the neutron energy group in the target coordinate direction u+.

[0076] 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:

[0077]

[0078] Wherein, Cf1 represents the first coefficient, h represents the node size, and D represents the diffusion coefficient. represents the coupling correction 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 g 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 coupling correction factor of the g neutron energy group in the target coordinate direction u-.

[0079] In step S202 of some embodiments, the second coefficient corresponding to the first average neutron flux density of the first 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.

[0080] In one implementation, the expression of the second coefficient may be as follows:

[0081]

[0082] in, Cf2 represents 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.

[0083] In step S203 of some embodiments, a neutron integral balance equation is constructed by taking the sum of the first product and each second product as 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 at the first moment, and the second product is the product of the second coefficient and the corresponding second average neutron flux density at the first moment.

[0084] In one embodiment, the expression of the above neutron integral balance equation can be expressed as follows (6):

[0085]

[0086] 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 average neutron source term within the block for the neutron energy group of the first block, t n+1 Indicates that the current running time is the first time.

[0087] The calculation formula of the average neutron source term in the above node can be expressed as follows:

[0088]

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

[0090] In some embodiments, step S301 may include but is not limited to steps S401 to S404.

[0091] Step S401, 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.

[0092] Step S402, 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.

[0093] Step S403: determining a coupling correction 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.

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

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

[0096] In step S401 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 (8):

[0097]

[0098] 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.

[0099] In step S402 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 (9):

[0100]

[0101] 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.

[0102] In step S403 of some embodiments, the coupling correction factor of the first node in the target coordinate direction is expressed as follows (10):

[0103]

[0104] in, Represents the coupling correction factor of the first node in the target coordinate direction.

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

[0106] Step S501: Determine first cross-sectional data of a neutron macroscopic removal cross-sectional area of ​​a first neutron energy group at the first moment according to target thermal parameters at the first moment.

[0107] Step S502, obtaining the coupling correction factor of the first neutron energy group in each coordinate direction at the first moment, the target parameter and the average neutron source term in the node.

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

[0109] 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.

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

[0111]

[0112] 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.

[0113] Then, the cross-sectional data at the first moment calculated by the above formula (11) 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 (11) are updated according to the rod positions of each group of control rods by the following formula (12):

[0114]

[0115] 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.

[0116] It should be noted that if the cross-section type is the macroscopic absorption cross-section of the fission products of the neutron energy group, the corrected macroscopic absorption cross-section of the fission products of the neutron energy group is calculated according to equations (11) and (12), and then the cross-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 previous moment by the following equation (13):

[0117]

[0118] 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 previous moment; N Xe 、N Sm The densities are 135Xe and 149Sm respectively.

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

[0120] 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, and the first target node and the first neutron energy group are used as objects to illustrate step S501-step S502. Specifically, according to the above-mentioned cross-sectional data calculation formulas (11) and (12), the first cross-sectional data of the neutron macroscopic removal cross-sectional area of ​​the first neutron energy group of the first target node is determined according to the target thermal parameters.

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

[0122] In step S503 of some embodiments, the first cross-sectional data of the first neutron energy group at the first moment, the coupling correction factors of the first neutron energy group in each coordinate direction at the first moment, 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 at the current moment 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 at the current moment.

[0123] In some embodiments, the core transient simulation result includes the delayed neutron precursor nuclear density of the first target node at the first moment, and step S103 may include but is not limited to steps S601 to S603.

[0124] Step S601: for each neutron energy group of the first target node, based on the target thermal parameters at a second moment and the neutron integral balance equation of the core of the nuclear reactor, determine a first average neutron flux density of the neutron energy group of the first target node at the second moment, the second moment being a moment before the first moment.

[0125] Step S602: 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 at the second moment according to the target thermal parameters at the second moment; and determine the first target value of the neutron energy group according to the second cross-sectional data of the neutron energy group and the first average neutron flux density of the neutron energy group of the first target node at the second moment.

[0126] Step S603, determining the delayed neutron precursor nuclear density of the first target node at the first moment according to the first target value of each of the neutron energy groups and the delayed neutron precursor nuclear density of the first target node at the second moment.

[0127] This embodiment provides a method for calculating the density of each group of delayed neutron precursor nuclei, and the calculation formula can be expressed as follows (14):

[0128]

[0129] 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′ neutron energy group, represents the average neutron flux density of the g′ neutron energy group in the ijk block, 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, a 1m and a 2m Represents the coefficient.

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

[0131]

[0132] Among them, Δt n Indicates the size of a time step (i.e. the time difference between the second moment and the first moment).

[0133] 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.

[0134] In this embodiment, Recorded as the first target value.

[0135] In step S601 of some embodiments, first, the first average neutron flux density of each neutron energy group of the first target node at the moment before the current moment (ie, the second moment) is calculated according to the above steps S501 to S503.

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

[0137] In step S603 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. According to the above formula (14), 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, the first target value of each neutron energy group corresponding to the first target node, the first average neutron flux density of the neutron energy group of the first target node at the second time, and the first coefficient a 1m and the second coefficient a 2m , calculate the delayed neutron precursor nuclear density of the first target block.

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

[0139] Step S701, obtaining the concentration of fission products of a first target node at a second moment, where the second moment is a moment before the first moment;

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

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

[0142] Step S704, obtaining the concentration change of the fission product from the second moment to the first moment, wherein the concentration change is determined by target information, the target information includes relevant parameters of the fission product, and the concentration of the fission product of the first target node at the second moment, the relevant parameters of the fission product include the fission yield and decay constant of the fission product; if the fission product is 135I or 149Pm, the target information also includes a second target value of the fission product of each neutron energy group of the first target node at the first moment; if the fission product is 135Xe, the target information also includes a third target value, a fourth target value and the second target value of the fission product of each neutron energy group of the first target node at the first moment, the third target value The fourth target value is determined according to the decay constant of 135I and the concentration of 135I at the first moment, and the fourth target value is determined according to the cross-sectional data of the microscopic absorption cross-section of 135Xe of each neutron energy group of the first target node at the first moment, and the first average neutron flux density of each neutron energy group at the first moment; if the fission product is 149Sm, the target information also includes a fifth target value and a sixth target value, the fifth target value is determined according to the decay constant of 149Pm and the concentration of 149Pm at the first moment, and the sixth target value is determined according to the cross-sectional data of the microscopic absorption cross-section of 149Sm of each neutron energy group of the first target node at the first moment, and the first average neutron flux density of each neutron energy group at the first moment.

[0143] Step S705: Determine the concentration of the fission products of the first target node at the first moment according to the concentration of the fission products of the first target node at the second moment and the concentration change of the fission products from the second moment to the first moment.

[0144] In this embodiment, a calculation formula for the fission product concentration of each node is provided, wherein the calculation formula for the fission product 135I concentration at the first moment is shown in the following formula (16):

[0145]

[0146] in, Indicates the fission product 135I concentration of the node, 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 ijk block g′ group, represents the average neutron flux density of the ijk node g′ group, t n represents the moment before the current moment (i.e., the second moment), t n+1 represents the current time, Δtn Represents a time step (i.e., from the second moment to the first moment).

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

[0148]

[0149] in, represents the fission product 135Xe concentration of the ijk node, γ I represents the fission yield of fission product 135I, γ Xe represents 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 g′ neutron energy group in the ijk block, represents the average neutron flux density of the neutron energy group of the ijk block g′, It represents the microscopic absorption cross section of the fission product 135Xe of the g′ neutron energy group in the ijk node at the first moment.

[0150] The calculation formula of the fission product 149Pm concentration at the first moment is shown in the following formula (18):

[0151]

[0152] in, represents the fission product 149Pm concentration of the ijk 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 g′ neutron energy group in the ijk block, Represents the average neutron flux density of the g′ neutron energy group in the ijk node.

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

[0154]

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

[0156] In this embodiment, Recorded as the second target value, Recorded as the third target value, Recorded as the fourth target value, Recorded as the fifth target value, Recorded as the sixth target value.

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

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

[0159] For the concentration calculation formula of the fission product 149Sm, the concentration change of the fission product from the second moment to the first moment is determined based on the relevant parameters of 149Sm, the concentration of 149Sm of the first target node at the second moment, the decay constant of 149Pm, the concentration of 149Pm at the first moment, the cross-sectional data of the microscopic absorption cross-section of 149Sm of each neutron energy group of the first target node at the first moment, and the first average neutron flux density of each neutron energy group at the first moment.

[0160] In step S701 of some embodiments, the fission product concentration at a moment before the current moment is obtained.

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

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

[0163] In step S704 of some embodiments, the concentration change of the fission product from the second moment to the first moment is calculated.

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

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

[0166] Step S801, for each neutron energy group of the first target node, based on the target thermal parameters at the first moment 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 at the first moment.

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

[0168] Step S803: determining the fission power of the first target node at the first moment according to the seventh target value of each neutron energy group of the first target node.

[0169] 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 (20):

[0170]

[0171] in, represents the fission power of the g′ neutron energy group in the ijk block, represents the macroscopic energy generation cross section of the g′ neutron energy group in the ijk block, represents the average neutron flux density of the g′ neutron energy group in the ijk block, V ijk represents the volume of the ijk node, t n+1 Indicates that the current moment is the first moment.

[0172] In this embodiment, Recorded as the seventh target value.

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

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

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

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

[0177] The decay heat of the first target node at the first moment is acquired, and the nuclear power of the first target node at the first moment is obtained according to the decay heat of the first target node at the first moment and the fission power of the first target node at the first moment.

[0178] This embodiment provides a calculation formula for decay heat of a node, which is specifically shown in the following formula (21):

[0179]

[0180] Among them, P d 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 239Decay heat contribution of U, P Np express 239 Decay heat contribution of Np, t n+1 Indicates that the current moment is the first moment.

[0181] This embodiment also provides a calculation formula for the nuclear power of a node, which is specifically shown in the following formula (22):

[0182]

[0183] Among them, P ujk represents the core power of the ijk node, represents the fission power of the ijk node, Represents the decay heat of the ijk node.

[0184] In step S802 of some embodiments, the decay heat of the first target node at the first moment 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 at the first moment is calculated to obtain the nuclear power of the first target node at the first moment.

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

[0186] Among them, by calculating the sum of the nuclear powers of each node in the core at the first moment, the total nuclear power of the core at the first moment is obtained. f It can be calculated by the following formula (23):

[0187]

[0188] By calculating the sum of the decay heat of each node in the core at the first moment, the total decay heat of the core at the first moment is obtained. d It can be calculated by the following formula (24):

[0189]

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

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

[0192] In some embodiments, the core transient simulation result includes the input value of the ex-core neutron detector, and the input value of the ex-core neutron detector is used to simulate the result of the core neutron leakage to the ex-core neutron detector. Step S103 may include but is not limited to steps S901 to S903:

[0193] Step S901, for each neutron energy group of the measurement node measured by the ex-core neutron detector, based on the target thermal parameters at the first moment 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 measurement node at the first moment.

[0194] Step S902, for each measurement node measured by the ex-core neutron detector, determine the eighth target value of the first partition according to the thermal group neutron flux density of the measurement node corresponding to the first partition of the ex-core neutron detector at the first moment, the sensitivity coefficient of the ex-core neutron detector to the measurement node, and the length of the first partition; the first partition is any partition of the ex-core neutron detector; according to the eighth target value of each partition, determine the ninth target value of the measurement node.

[0195] Step S903: determining the input value of the ex-core neutron detector at the first moment according to the ninth target value of each measurement node.

[0196] This embodiment also provides a calculation formula for the input value of the out-of-core neutron detector, which is specifically shown in the following formula (26):

[0197]

[0198] In the formula, A dect Represents the input value of the neutron detector; I represents the number of neutron detector partitions; N represents the number of measurable nodes of the neutron detector; 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 of the detector corresponding to the measurement node n in the i-th partition; l i represents the length of the neutron detector partition i, t n+1 Indicates that the current moment is the first moment.

[0199] In this embodiment, Recorded as the eighth target value, Recorded as the ninth target value.

[0200] The measurement node is a node that can be measured by the ex-core neutron detector, and the first partition is any partition among the partitions of the neutron detector.

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

[0202] In step S902 of some embodiments, for each measurement node measured by the ex-core neutron detector, the product of the thermal group neutron flux density of the measurement node corresponding to the first partition of the ex-core neutron detector at the first moment, the sensitivity coefficient of the ex-core neutron detector to the measurement node, and the length of the first partition is calculated to determine the eighth target value of the first partition. And calculate the sum of the eighth target values ​​of each partition to get the ninth target value of the measurement node

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

[0204] In 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 transient simulation result is calculated through the average neutron flux density, thereby obtaining a more accurate core transient simulation result, thereby improving the effectiveness of the core transient simulation.

[0205] 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.

[0206] For ease of understanding, a specific embodiment is used as an example for illustration:

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

[0208] Step 2: The core transient model reads the model input card, completes the initialization of the core transient model, and establishes the core simulation model of the nuclear reactor.

[0209] Step 3: The core transient 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, transient calculations are performed at the current moment.

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

[0211] Step 5: Define the dynamic frequency expression and use the implicit difference method to differentiate the time term in the dynamic frequency expression to obtain the differential expression of the dynamic frequency. Substitute the differential expression into the neutron kinematic equation to obtain the transient neutron diffusion equation. Use the semi-analytical nodal method to spatially discretize the transient neutron diffusion equation, perform volume integration within the nodal block, and introduce the coupling correction factor to obtain the expression of the neutron integral balance equation for the entire core (6).

[0212] The average neutron flux density of the node at the moment before the current moment is given, and the coupling correction factor at the current moment is calculated. The average neutron flux density of the neutron energy group of the node at the current moment is calculated by the above formula (6). Specifically, the average neutron source term in the node is calculated according to formula (7), and the net neutron flow and its correction flow are solved by using the transverse integration method and the residual weight method, and then the coupling correction factor at the current moment is solved. Specifically, the coupling correction factor of the neutron energy group at the current moment in each coordinate direction is calculated according to formulas (8), (9) and (10).

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

[0214] The delayed neutron precursor nuclear density is calculated according to the above formula (14).

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

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

[0217] According to the above formulas (23), (24) and (25), the total fission power, decay heat and nuclear power of the core are calculated respectively.

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

[0219] Step 9: Synchronously update and modify thermal parameters, control instructions and other parameters online through the simulation platform.

[0220] 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.

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

[0222] Figure 2 This is a schematic diagram of the structure of the nuclear reactor core transient simulation device provided in the embodiment of the present application. Figure 2 The embodiment of the present application also provides a nuclear reactor core transient simulation device 800, 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 can implement the above-mentioned nuclear reactor core transient simulation method. The nuclear reactor core transient simulation device 800 includes:

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

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

[0225] The transient simulation module 803 is used to respond to the transient calculation instruction through the core simulation model, and determine the transient 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 transient neutron diffusion equation by a nonlinear semi-analytical block method, the transient neutron diffusion equation is calculated by an implicit difference method, and the average neutron source term within the block in the neutron integral balance equation is determined by volume integration within the block.

[0226] The specific implementation of the nuclear reactor core transient simulation device can refer to the specific embodiment of the nuclear reactor core transient simulation method mentioned above, which will not be repeated here.

[0227] 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 transient simulation method when executing the computer program. The electronic device can be any intelligent terminal including a tablet computer, a car computer, etc.

[0228] See also Figure 3 , Figure 3 The hardware structure of an electronic device of another embodiment is illustrated, and the electronic device includes:

[0229] 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;

[0230] The memory 102 can 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 can store an operating system and other application programs. When the technical solution provided in the embodiment 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 core transient simulation method of the nuclear reactor of the embodiment of the present application;

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

[0232] 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.);

[0233] 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);

[0234] 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 .

[0235] 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 transient simulation method.

[0236] 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.

[0237] 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.

[0238] 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.

[0239] 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.

[0240] 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.

[0241] 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.

[0242] 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.

[0243] 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.

[0244] 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.

[0245] 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.

[0246] 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.

[0247] 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 transient 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 transient 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 transient calculation instruction and determines the transient 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 transient neutron diffusion equation by a nonlinear semi-analytical block method, the transient neutron diffusion equation is calculated by an implicit difference 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 Before the neutron integral balance equation based on the target thermal parameters and the core of the nuclear reactor, the method further includes: The time derivative of the neutron flux density is taken as the first parameter; expressing a ratio of the first parameter to the neutron flux density as a first dynamic frequency; The time derivative of the delayed neutron precursor nuclear density is taken as the second parameter; expressing the ratio of the second parameter to the neutron precursor nuclear density as a second dynamic frequency; The first dynamic frequency and the second dynamic frequency are respectively discretized by implicit difference method to obtain a differential representation of the first dynamic frequency and a differential representation of the second dynamic frequency; Obtaining the transient neutron diffusion equation according to the differential representation of the first dynamic frequency and the differential representation of the second dynamic frequency; The transient neutron diffusion equation is spatially discretized to obtain the neutron integral balance equation.

3. 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 at the first moment, the second average neutron flux density of the neutron energy group of the second node at the first moment, 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 coupling correction 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 at the first moment, the second average neutron flux density at the first moment, and the average neutron source term within the node.

4. The method according to claim 3, 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 coupling correction 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 coupling correction 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.

5. The method according to claim 3, characterized in that: The core transient simulation result includes a first average neutron flux density of a first neutron energy group of a first target node at a first moment, wherein the first target node is any node in the core of the nuclear reactor, the first target node includes a plurality of neutron energy groups, and the first neutron energy group is any neutron energy group of the first target node; The step of determining the transient 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 comprises: Determining first cross-sectional data of a neutron macroscopic removal cross-sectional area of ​​a first neutron energy group at the first moment according to the target thermal parameters at the first moment; Obtaining the coupling correction factor of the first neutron energy group in each coordinate direction at the first moment, the target parameter and the average neutron source term in the node; The first cross-sectional data of the first neutron energy group at the first moment, the coupling correction factors of the first neutron energy group in each coordinate direction at the first moment, 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 at the first moment.

6. The method according to claim 1, characterized in that The core transient simulation result includes the delayed neutron precursor nuclear density of the first target node at the first moment, wherein the first target node is any node in the core of the nuclear reactor, and the first target node includes a plurality of neutron energy groups; The step of determining the transient 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 comprises: For each neutron energy group of the first target node, based on the target thermal parameters at a second moment 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 at the second moment, the second moment being a moment before the first moment; 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 at the second moment are determined according to the target thermal parameters at the second moment; 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 of the neutron energy group of the first target node at the second moment; The delayed neutron precursor nuclear density of the first target node at the first moment is determined according to the first target value of each of the neutron energy groups and the delayed neutron precursor nuclear density of the first target node at the second moment.

7. The method according to claim 1, characterized in that The core transient simulation result includes the concentration of fission products of the first target node at the first moment, wherein the first target node is any node in the core of the nuclear reactor, the first target node includes a plurality of neutron energy groups, and the fission products include at least one of fission product 135I, fission product 135Xe, fission product 149Pm and fission product 149Sm; The step of determining the transient 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 comprises: Acquire the concentration of the fission product of the first target node at a second moment, where the second moment is a moment before the first moment; For each neutron energy group of the first target node, based on the target thermal parameters at the first moment and the neutron integral balance equation of the nuclear reactor, determining a first average neutron flux density of the neutron energy group of the first target node at the first moment; For each neutron energy group of the first target node, according to the target thermal parameters at the first moment, the cross-sectional data of the macroscopic absorption cross section of the fission products of the neutron energy group are determined, and according to the cross-sectional data of the macroscopic absorption cross section of the fission products of the neutron energy group and the first average neutron flux density of the neutron energy group at the first moment, the second target value of the fission products of the neutron energy group at the first moment is determined; Obtaining a concentration change of the fission product from the second moment to the first moment, wherein the concentration change is determined by target information, the target information including relevant parameters of the fission product and the concentration of the fission product of the first target node at the second moment, the relevant parameters of the fission product including the fission yield and decay constant of the fission product; if the fission product is 135I or 149Pm, the target information also includes a second target value of the fission product of each neutron energy group of the first target node at the first moment; if the fission product is 135Xe, the target information also includes a third target value, a fourth target value and a second target value of the fission product of each neutron energy group of the first target node at the first moment, the third target value being determined according to the The target information is determined according to the decay constant of 135I and the concentration of 135I at the first moment, and the fourth target value is determined according to the cross-sectional data of the microscopic absorption cross-section of 135Xe of each neutron energy group of the first target node at the first moment, and the first average neutron flux density of each neutron energy group at the first moment; if the fission product is 149Sm, the target information also includes a fifth target value and a sixth target value, the fifth target value is determined according to the decay constant of 149Pm and the concentration of 149Pm at the first moment, and the sixth target value is determined according to the cross-sectional data of the microscopic absorption cross-section of 149Sm of each neutron energy group of the first target node at the first moment, and the first average neutron flux density of each neutron energy group at the first moment; The concentration of the fission product of the first target node at the first moment is determined according to the concentration of the fission product of the first target node at the second moment and the concentration change of the fission product from the second moment to the first moment.

8. The method according to claim 1, characterized in that The core transient simulation result includes the fission power of a first target node, wherein the first target node is any node in the core of the nuclear reactor, and the first target node includes a plurality of neutron energy groups; The step of determining the transient 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 comprises: For each neutron energy group of the first target node, based on the target thermal parameters at the first moment 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 at the first moment; For each neutron energy group of the first target node, according to the target thermal parameters at the first moment, the cross-sectional data of the macroscopic energy generation cross-section of the neutron energy group is determined; and according to the cross-sectional data of the macroscopic energy generation cross-section of the neutron energy group, the first average neutron flux density of the neutron energy group of the first target node at the first moment, and the volume of the first target node, the seventh target value of the neutron energy group is determined; determining the fission power of the first target node at the first moment according to the seventh target value of each neutron energy group of the first target node; The core transient simulation result includes the nuclear power of the first target node at the first moment, and after determining the fission power of the first target node according to the seventh target value of each neutron energy group of the first target node, the method further includes: The decay heat of the first target node at the first moment is acquired, and the nuclear power of the first target node at the first moment is obtained according to the decay heat of the first target node at the first moment and the fission power of the first target node at the first moment.

9. The method according to claim 1, characterized in that: The core transient simulation result includes an input value of an external neutron detector, and the determining of the core transient 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 includes: For each neutron energy group of the measurement node measured by the ex-core neutron detector, based on the target thermal parameters at the first moment and the neutron integral balance equation of the core of the nuclear reactor, determine a first average neutron flux density of the neutron energy group of the measurement node at the first moment; For each measurement node measured by the ex-core neutron detector, determine the eighth target value of the first partition according to the thermal group neutron flux density of the measurement node corresponding to the first partition of the ex-core neutron detector at the first moment, the sensitivity coefficient of the ex-core neutron detector to the measurement node, and the length of the first partition; the first partition is any partition of the various partitions of the ex-core neutron detector; determine the ninth target value of the measurement node according to the eighth target value of each partition; The input value of the ex-core neutron detector at the first moment is determined according to the ninth target value of each measurement node.

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

11. A computer-readable storage medium storing a computer program, characterized in that: When the computer program is executed by a processor, the method for simulating the core transient state of a nuclear reactor according to any one of claims 1 to 9 is implemented.

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