Modeling method and system of non-uniform reactor core three-dimensional neutron physical model, storage medium and electronic equipment

Abstract

The invention relates to a modeling method and system for a non-uniform reactor core three-dimensional neutron physical model, a storage medium and electronic equipment, and the method comprises the steps: carrying out the material component analysis of a fuel assembly of a reactor, and obtaining a plurality of different segments; performing homogenization treatment on each of the plurality of different segments to obtain a plurality of different uniform segments; performing calculation based on the components in the uniform segment and the thermal parameters to obtain section parameters of the uniform segment; modeling based on the section parameters and in combination with the geometric parameters of the reactor core to obtain a multi-group neutron diffusion equation; solving the multi-group neutron diffusion equation to obtain the neutron flux of uniform segments; and calculating according to the neutron flux in combination with the fission cross section to obtain the three-dimensional power distribution of the reactor core. The method can be suitable for modeling of the reactor core composed of non-uniform components, and the model precision is high.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear energy neutron physics modeling, and more specifically, to a modeling method, system, storage medium and electronic equipment for a three-dimensional neutron physics model of a non-uniform core. Background Art

[0002] As one of the important models of nuclear reactors, the three-dimensional neutron physics model is also indispensable in application scenarios that require real-time simulation. The accuracy and real-time performance of its simulation directly affect the application performance of the nuclear reactor model. Traditional nuclear reactors usually use nuclear fuel with the same enrichment in the axial direction, and the burnable poison rods are evenly arranged. However, in recent years, more and more new reactors have adopted axial non-uniform fuel arrangement in design, and the burnable poison rods are also arranged in a non-uniform manner. Therefore, the reactor is non-uniform in the axial direction, and at the same time, the radial direction is also the same as the traditional nuclear reactor core design, and the enrichment of fuel assemblies and the number of burnable poison rods at different positions are inconsistent. In the field of traditional real-time three-dimensional neutron physics model simulation, the same component is usually homogenized and homogenized inside the component, but this method encounters difficulties when applied to the core composed of such axial non-uniform components, resulting in insufficient accuracy of the model. Summary of the invention

[0003] The technical problem to be solved by the present invention is to provide a modeling method, system, storage medium and electronic equipment for a three-dimensional neutron physics model of a non-uniform core in view of the problems existing in the prior art.

[0004] The technical solution adopted by the present invention to solve the technical problem is: constructing a modeling method of a three-dimensional neutron physics model of a non-uniform core, comprising the following steps:

[0005] Conduct material composition analysis on the reactor fuel assembly to obtain multiple different nodes;

[0006] Performing a homogenization process on each of the plurality of different nodes to obtain a plurality of different uniform nodes;

[0007] The following operations are performed on each of the plurality of different uniform blocks:

[0008] Calculating based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node;

[0009] Modeling is performed based on the cross-sectional parameters and in combination with core geometric parameters to obtain a multi-group neutron diffusion equation;

[0010] Solving the multi-group neutron diffusion equation to obtain the neutron flux of the uniform node;

[0011] The three-dimensional power distribution of the core is obtained by calculating according to the neutron flux in combination with the fission cross section.

[0012] In the modeling method of the non-uniform core three-dimensional neutron physics model of the present invention, the material composition analysis of the fuel assembly of the reactor to obtain a plurality of different nodes includes:

[0013] Material composition analysis is performed on the fuel assembly in radial and axial directions to obtain the multiple different nodes.

[0014] In the modeling method of the non-uniform core three-dimensional neutron physics model of the present invention, the material composition analysis of the fuel assembly in radial and axial directions to obtain the multiple different nodes includes:

[0015] Dividing the fuel assembly into blocks in radial direction with the edge of the fuel assembly as the boundary;

[0016] In the axial direction, the nodes are divided based on the positions where the enrichment changes and the number of the combustible poison rods changes as boundaries to obtain the multiple different nodes.

[0017] In the modeling method of the three-dimensional neutron physics model of the non-uniform core of the present invention, the calculation based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node includes:

[0018] Calculating based on the components and thermal parameters in the uniform node to obtain initial parameters;

[0019] determining a control rod model of the uniform node;

[0020] The initial parameters are updated based on the control rod model.

[0021] In the modeling method of the three-dimensional neutron physics model of the non-uniform core of the present invention, the calculation based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node also includes:

[0022] determining a fission product model of the uniform node;

[0023] The initial parameters are updated based on a fission product model to obtain cross-sectional parameters of the uniform node.

[0024] The present invention also provides a modeling system for a three-dimensional neutron physics model of a non-uniform core, comprising:

[0025] A node division unit is used to analyze the material composition of the fuel assembly of the reactor to obtain a plurality of different nodes;

[0026] A homogenization processing unit, used for performing homogenization processing on each of the plurality of different nodes to obtain a plurality of different uniform nodes;

[0027] A three-dimensional modeling unit, used for performing calculations based on the components and thermal parameters in the uniform node to obtain cross-sectional parameters of the uniform node;

[0028] Modeling is performed based on the cross-sectional parameters and in combination with core geometric parameters to obtain a multi-group neutron diffusion equation;

[0029] Solving the multi-group neutron diffusion equation to obtain the neutron flux of the uniform node;

[0030] The three-dimensional power distribution of the core is obtained by calculating according to the neutron flux in combination with the fission cross section.

[0031] In the modeling system of the non-uniform core three-dimensional neutron physics model of the present invention, the node division unit includes: a radial division module and an axial division module;

[0032] The radial division module is used to divide the fuel assembly into blocks in the radial direction with the edge of the fuel assembly as the boundary;

[0033] The axial division module is used to divide the nodes in the axial direction using the positions where the enrichment changes and the number of combustible poison rods changes as boundaries to obtain the multiple different nodes.

[0034] In the modeling system of the non-uniform core three-dimensional neutron physics model of the present invention, the three-dimensional modeling unit includes: an equation building module, a neutron flux calculation module and a three-dimensional power calculation module;

[0035] The equation building module is used to build a multi-group neutron diffusion equation based on the cross-sectional parameters and in combination with core geometric parameters;

[0036] The neutron flux calculation module is used to solve the multi-group neutron diffusion equation to obtain the neutron flux of the uniform block;

[0037] The three-dimensional power calculation module performs calculations based on the neutron flux and in combination with the fission cross section to obtain the three-dimensional power distribution of the core.

[0038] The present invention also provides a storage medium storing a computer program, wherein the computer program is suitable for being loaded by a processor to execute the steps of the modeling method of the three-dimensional neutron physics model of the non-uniform core as described above.

[0039] The present invention also provides an electronic device, comprising a memory and a processor, wherein the memory stores a computer program, and the processor executes the steps of the modeling method of the non-uniform core three-dimensional neutron physics model as described above by calling the computer program stored in the memory.

[0040] The modeling method, system, storage medium and electronic device of the three-dimensional neutron physics model of the non-uniform core of the present invention have the following beneficial effects: including: performing material composition analysis on the fuel assembly of the reactor to obtain multiple different nodes; performing homogenization treatment on each of the multiple different nodes to obtain multiple different uniform nodes; performing the following operations on each uniform node of the multiple different uniform nodes: calculating based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node; establishing a multi-group neutron diffusion equation based on the cross-sectional parameters and in combination with the core geometric parameters; solving the multi-group neutron diffusion equation to obtain the neutron flux of the uniform node; calculating based on the neutron flux and in combination with the fission cross section to obtain the three-dimensional power distribution of the core. The present invention can be applied to the core modeling composed of non-uniform components, and the model has high accuracy. BRIEF DESCRIPTION OF THE DRAWINGS

[0041] The present invention will be further described below with reference to the accompanying drawings and embodiments, in which:

[0042] Figure 1 It is a schematic flow chart of a modeling method of a three-dimensional neutron physics model of a non-uniform core provided by the present invention;

[0043] Figure 2 This is a schematic diagram of dividing the fuel assembly provided by the present invention into sections;

[0044] Figure 3 It is a logic block diagram of a modeling system of a three-dimensional neutron physics model of a non-uniform core provided by the present invention. DETAILED DESCRIPTION

[0045] The following will be combined with the drawings in the embodiments of the present invention to clearly and completely describe the technical solutions in the embodiments of the present invention. Obviously, the described embodiments are only part of the embodiments of the present invention, not all of the embodiments. Based on the embodiments of the present invention, all other embodiments obtained by ordinary technicians in this field without creative work are within the scope of protection of the present invention.

[0046] The present invention provides a modeling method for a three-dimensional neutron physics model of a non-uniform core, so as to solve the difficulty of modeling non-uniform components and improve the accuracy of the model.

[0047] Specifically, Figure 1 As shown, the modeling method of the non-uniform core three-dimensional neutron physics model includes the following steps:

[0048] Step S101: Analyze the material composition of the fuel assembly of the reactor to obtain a plurality of different nodes.

[0049] Specifically, it should be noted that in the embodiment of the present invention, the reactor core is composed of a plurality of fuel assemblies.

[0050] In an embodiment of the present invention, performing material composition analysis on a fuel assembly of a reactor to obtain a plurality of different nodes includes: performing material composition analysis on the fuel assembly in radial and axial directions to obtain a plurality of different nodes.

[0051] In an embodiment of the present invention, material composition analysis is performed on the fuel assembly in radial and axial directions to obtain a plurality of different nodes, including: dividing the nodes in radial direction using the edge of the fuel assembly as the boundary; dividing the nodes in axial direction using the position where the enrichment changes and the number of combustible poison rods changes as the boundary to obtain a plurality of different nodes.

[0052] Specifically, material component analysis is performed in radial and axial directions, and the nodes are divided based on the edge of each component in radial direction, and based on the position where the enrichment changes and the number of burnable poison rods changes in axial direction, so as to ensure that the fuel enrichment and the number of burnable poison rods in the same three-dimensional node of the neutron physics model will not change in space. Figure 2 The fuel assembly shown is divided into blocks at the boundaries of the burnable poison rods to ensure the consistency of the burnable poison rods within the blocks.

[0053] Step S102: performing homogenization processing on each of the multiple different nodes to obtain multiple different uniform nodes.

[0054] In an embodiment of the present invention, by homogenizing the divided nodes within the nodes, the components at each position in each node can be the same and the physical properties are also the same, that is, the pressure, temperature, macroscopic cross-section and other parameters at each position in each node are constant, and the values ​​at each position in the node are the same.

[0055] Step S103: Calculate based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node.

[0056] In the embodiment of the present invention, the cross-sectional parameters of the uniform node can be calculated by the following formula:

[0057] ∑ xg =f(T f ,T M ,ρ M ,C B ) (1);

[0058] Among them, Σ xg represents the macroscopic cross section of the nuclear reaction x of neutrons with energy group g, T f represents the fuel temperature, T M represents the coolant temperature, ρ M represents the moderator density, C B Indicates the boron concentration.

[0059] For the same node, it has multiple types of cross-sections, and generally speaking, for neutron physics models, neutron groups are divided according to different neutron energies, but these can be attributed to different cross-section types. For a specific cross-section of a node, Formula 1 is not exactly the same. The same polynomial can be used, but its coefficients are different; or different polynomials can be used for different cross-sections to perform cross-section fitting calculations. In short, through one or more sets of formulas, all cross-section parameters of each node under a specific thermal parameter can be obtained.

[0060] In an embodiment of the present invention, calculations are performed based on the components and thermal parameters within the uniform node to obtain cross-sectional parameters of the uniform node, including: calculating based on the components and thermal parameters within the uniform node to obtain initial parameters; determining a control rod model of the uniform node; and updating the initial parameters based on the control rod model.

[0061] Specifically, in the neutron physics model, the influence of the control rod model needs to be considered when calculating the cross-sectional parameters of the uniform node. The influence of the control rod model is updated by the following formula:

[0062]

[0063] (2) Corrected cross section; Σ n,NR,X,g : The cross section of the node without control rod inserted; ΔΣ n,X,g : The cross section after the segment is fully inserted into the control rod; f n,R : A coefficient based on the insertion scale.

[0064] Furthermore, in an embodiment of the present invention, calculation based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node also includes: determining a fission product model of the uniform node; and updating the initial parameters based on the fission product model to obtain the cross-sectional parameters of the uniform node.

[0065] Specifically, the influence of fission products, especially xenon (Xe-135) and samarium (Sm-149), needs to be considered in the neutron physics model. Therefore, the calculation of the cross section needs to introduce a fission product-related model to characterize the influence of fission products on the cross section. The influence of fission products can be updated by the following formula:

[0066]

[0067] (3) In the formula, Σ ag : absorption cross section of energy group g; Increased absorption cross section due to Xe-135; The increase in absorption cross section due to Sm-149.

[0068]

[0069] N Xe , N Sm Respectively represent 135 Xe and 149 The density of Sm, and It is its microscopic cross-section.

[0070] Step S104: Modeling is performed based on the cross-sectional parameters and in combination with the core geometric parameters to obtain a multi-group neutron diffusion equation.

[0071] In the embodiment of the present invention, the cross-sectional parameters obtained in step S103 are combined with the geometric parameters of the core (ie, the length, width, and height of the node) to perform modeling, thereby obtaining a multi-group neutron diffusion equation.

[0072] Specifically, for the volume: The agglomeration m, ( The distance between the center and the boundary of the hexagonal node is the same as the distance between the center and the boundary of the hexagonal node. Here is a hexagonal node, and Vm represents the volume. The multi-group neutron diffusion equation is as follows:

[0073]

[0074] (4) In the formula, The diffusion coefficient of the node m and the neutron energy group g; Node m, neutron flux of neutron energy group g; Node m, total consumption cross section for neutron energy group g; Block m, the fission reaction cross section when the neutron energy group is not g; Node m, the reaction cross section of neutrons transforming from other energy groups to energy group g; In block m, the neutron source intensity of neutrons in energy group g; xg represents the amount of neutrons produced by energy group g after each fission; k eff : represents the effective value-added coefficient.

[0075] in, is the object to be solved, and These are the cross-sectional parameters of each node we obtained above.

[0076] Step S105: Solve the multi-group neutron diffusion equation to obtain the neutron flux of the uniform node.

[0077] Specifically, in the embodiment of the present invention, the neutron flux of each uniform node is obtained by solving equation (4) in step S104.

[0078] Step S106: Calculate according to the neutron flux and in combination with the fission cross section to obtain the three-dimensional power distribution of the core.

[0079] In an embodiment of the present invention, the three-dimensional power distribution of the core is obtained by calculating according to the neutron flux and in combination with the fission cross section, including: calculating the total power value of each uniform node according to the neutron flux and in combination with the fission cross section, and determining the three-dimensional power distribution of the core according to the total power value of all uniform nodes. That is, after obtaining the neutron flux of each uniform node through steps S103 to S105, the neutron flux of each uniform node is used in combination with its own fission cross section to calculate through step S106 to obtain the total power value of each uniform node, and then the three-dimensional power distribution of each fuel assembly can be determined, and finally the three-dimensional power distribution of the core can be determined based on the three-dimensional power distribution of the fuel assembly. Among them, the power of each position is equal to its flux multiplied by the fission cross section. The cross section in neutron physics is actually not an area, but the probability of such a nuclear reaction. Among them, the calculation formula for the total power value of each uniform node is:

[0080]

[0081] (5) In the formula, Neutron flux of uniform nodes; Fission cross section of a uniform node.

[0082] The present invention can establish a complete neutron physics model through the above process and method. By dividing the nodes and setting the cross-section calculation formula at different positions, the problem of different components caused by inconsistent fuel types at different positions can be avoided. Especially when facing different radial fuel types, it is only necessary to set different cross-section formulas or coefficients to perform cross-section calculation, so that the model (i.e., equation (4)) can be applied to neutron physics modeling of various non-uniform reactor types, and the model has high accuracy.

[0083] refer to Figure 3 The present invention also provides a modeling system for a three-dimensional neutron physics model of a non-uniform core. Figure 3 As shown, the modeling system of the non-uniform core three-dimensional neutron physics model includes:

[0084] The node division unit 301 is used to perform material composition analysis on the fuel assembly of the reactor to obtain a plurality of different nodes.

[0085] Among them, the node division unit includes: a radial division module and an axial division module; the radial division module is used to divide the nodes in the radial direction with the edge of the fuel assembly as the boundary; the axial division module is used to divide the nodes in the axial direction with the position where the enrichment changes and the number of combustible poison rods changes as the boundary, so as to obtain multiple different nodes.

[0086] The homogenization processing unit 302 is used to perform homogenization processing on each of the multiple different nodes to obtain multiple different uniform nodes.

[0087] The three-dimensional modeling unit 303 is used to perform calculations based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node; perform modeling based on the cross-sectional parameters and in combination with the core geometric parameters to obtain the multi-group neutron diffusion equation; solve the multi-group neutron diffusion equation to obtain the neutron flux of the uniform node; and perform calculations based on the neutron flux and in combination with the fission cross section to obtain the three-dimensional power distribution of the core.

[0088] The three-dimensional modeling unit 303 includes: an equation building module, a neutron flux calculation module and a three-dimensional power calculation module.

[0089] The equation building module is used to establish multi-group neutron diffusion equations based on cross-sectional parameters and combined with core geometric parameters; the neutron flux calculation module is used to solve the multi-group neutron diffusion equations to obtain the neutron flux of uniform nodes; the three-dimensional power calculation module calculates based on the neutron flux and combined with the fission cross section to obtain the three-dimensional power distribution of the core.

[0090] Specifically, the specific coordination operation process between the units in the modeling system of the non-uniform core three-dimensional neutron physics model here can refer to the modeling method of the non-uniform core three-dimensional neutron physics model mentioned above, which will not be repeated here.

[0091] In addition, an electronic device of the present invention includes a memory and a processor; the memory is used to store a computer program; the processor is used to execute the computer program to implement a modeling method of a three-dimensional neutron physics model of a non-uniform core as described in any of the above items. Specifically, according to an embodiment of the present invention, the process described with reference to the flowchart above can be implemented as a computer software program. For example, an embodiment of the present invention includes a computer program product, which includes a computer program carried on a computer-readable medium, and the computer program contains a program code for executing the method shown in the flowchart. In such an embodiment, the computer program can be downloaded and installed by an electronic device and executed to perform the above functions defined in the method of the embodiment of the present invention. The electronic device in the present invention can be a terminal such as a notebook, a desktop, a tablet computer, a smart phone, or a server.

[0092] In addition, a storage medium of the present invention stores a computer program thereon, and when the computer program is executed by a processor, the modeling method of the three-dimensional neutron physical model of the non-uniform core of any one of the above items is implemented. Specifically, it should be noted that the above storage medium of the present invention can be a computer-readable signal medium or a computer-readable storage medium or any combination of the above two. The computer-readable storage medium can be, for example, but not limited to, an electrical, magnetic, optical, electromagnetic, infrared, or semiconductor system, device or device, or any combination of the above. More specific examples of computer-readable storage media can include, but are not limited to: an electrical connection with one or more wires, a portable computer disk, a hard disk, a random access memory (RAM), a read-only memory (ROM), an erasable programmable read-only memory (EPROM or flash memory), an optical fiber, a portable compact disk read-only memory (CD-ROM), an optical storage device, a magnetic storage device, or any suitable combination of the above. In the present invention, a computer-readable storage medium can be any tangible medium containing or storing a program, which can be used by an instruction execution system, device or device or used in combination with it. In the present invention, a computer-readable signal medium can include a data signal propagated in a baseband or as part of a carrier, which carries a computer-readable program code. Such propagated data signals may take a variety of forms, including but not limited to electromagnetic signals, optical signals, or any suitable combination of the above. Computer readable signal media may also be any computer readable medium other than computer readable storage media, which may send, propagate, or transmit programs for use by or in conjunction with an instruction execution system, apparatus, or device. The program code contained on the computer readable medium may be transmitted using any suitable medium, including but not limited to: wires, optical cables, RF (radio frequency), etc., or any suitable combination of the above.

[0093] The computer-readable medium may be included in the electronic device, or may exist independently without being incorporated into the electronic device.

[0094] In this specification, each embodiment is described in a progressive manner, and each embodiment focuses on the differences from other embodiments. The same or similar parts between the embodiments can be referred to each other. For the device disclosed in the embodiment, since it corresponds to the method disclosed in the embodiment, the description is relatively simple, and the relevant parts can be referred to the method part.

[0095] Professionals may further appreciate that the units and algorithm steps of each example described in conjunction with the embodiments disclosed herein can be implemented in electronic hardware, computer software, or a combination of the two. In order to clearly illustrate the interchangeability of hardware and software, the composition and steps of each example have been generally described in the above description according to function. Whether these functions are performed in hardware or software depends on the specific application and design constraints of the technical solution. Professionals and technicians may use different methods to implement the described functions for each specific application, but such implementation should not be considered to be beyond the scope of the present invention.

[0096] The steps of the method or algorithm described in conjunction with the embodiments disclosed herein may be implemented directly using hardware, a software module executed by a processor, or a combination of the two. The software module may be placed in a random access memory (RAM), a memory, a read-only memory (ROM), an electrically programmable ROM, an electrically erasable programmable ROM, a register, a hard disk, a removable disk, a CD-ROM, or any other form of storage medium known in the art.

[0097] The above embodiments are only for illustrating the technical concept and features of the present invention, and their purpose is to enable people familiar with the technology to understand the content of the present invention and implement it accordingly, and they cannot limit the scope of protection of the present invention. All equivalent changes and modifications made to the scope of the claims of the present invention should fall within the scope of the claims of the present invention.

Claims

1. A modeling method for a three-dimensional neutron physics model of a non-uniform core, characterized in that: The following steps are involved: Conduct material composition analysis on the reactor fuel assembly to obtain multiple different nodes; Performing a homogenization process on each of the plurality of different nodes to obtain a plurality of different uniform nodes; Calculating based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node; Modeling is performed based on the cross-sectional parameters and in combination with core geometric parameters to obtain a multi-group neutron diffusion equation; Solving the multi-group neutron diffusion equation to obtain the neutron flux of the uniform node; The three-dimensional power distribution of the core is obtained by calculating according to the neutron flux in combination with the fission cross section.

2. The modeling method of the non-uniform core three-dimensional neutron physics model according to claim 1 is characterized in that: The material composition analysis of the fuel assembly of the reactor to obtain a plurality of different nodes includes: Material composition analysis is performed on the fuel assembly in radial and axial directions to obtain the multiple different nodes.

3. The modeling method of the non-uniform core three-dimensional neutron physics model according to claim 2 is characterized in that: The performing material composition analysis on the fuel assembly in radial and axial directions to obtain the plurality of different nodes comprises: Dividing the fuel assembly into blocks in radial direction with the edge of the fuel assembly as the boundary; In the axial direction, the nodes are divided based on the positions where the enrichment changes and the number of the combustible poison rods changes as boundaries to obtain the multiple different nodes.

4. The modeling method of the non-uniform core three-dimensional neutron physics model according to claim 1 is characterized in that: The step of calculating based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node includes: Calculating based on the components and thermal parameters in the uniform node to obtain initial parameters; determining a control rod model of the uniform node; The initial parameters are updated based on the control rod model.

5. The modeling method of the non-uniform core three-dimensional neutron physics model according to claim 4 is characterized in that: The step of calculating based on the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node also includes: determining a fission product model of the uniform node; The initial parameters are updated based on a fission product model to obtain cross-sectional parameters of the uniform node.

6. A modeling system for a three-dimensional neutron physics model of a non-uniform core, characterized in that: include: A node division unit is used to analyze the material composition of the fuel assembly of the reactor to obtain a plurality of different nodes; A homogenization processing unit, used for performing homogenization processing on each of the plurality of different nodes to obtain a plurality of different uniform nodes; A three-dimensional modeling unit, used for calculating the components and thermal parameters in the uniform node to obtain the cross-sectional parameters of the uniform node; Modeling is performed based on the cross-sectional parameters and in combination with core geometric parameters to obtain a multi-group neutron diffusion equation; Solving the multi-group neutron diffusion equation to obtain the neutron flux of the uniform node; The three-dimensional power distribution of the core is obtained by calculating according to the neutron flux in combination with the fission cross section.

7. The modeling system of the non-uniform core three-dimensional neutron physics model according to claim 6, characterized in that: The node division unit includes: a radial division module and an axial division module; The radial division module is used to divide the fuel assembly into blocks in the radial direction with the edge of the fuel assembly as the boundary; The axial division module is used to divide the nodes in the axial direction using the positions where the enrichment changes and the number of combustible poison rods changes as boundaries to obtain the multiple different nodes.

8. The modeling system of the non-uniform core three-dimensional neutron physics model according to claim 6, characterized in that: The three-dimensional modeling unit includes: an equation building module, a neutron flux calculation module and a three-dimensional power calculation module; The equation building module is used to build a multi-group neutron diffusion equation based on the cross-sectional parameters and in combination with core geometric parameters; The neutron flux calculation module is used to solve the multi-group neutron diffusion equation to obtain the neutron flux of the uniform block; The three-dimensional power calculation module performs calculations based on the neutron flux and in combination with the fission cross section to obtain the three-dimensional power distribution of the core.

9. A storage medium, characterized in that: The storage medium stores a computer program, and the computer program is suitable for being loaded by a processor to execute the steps of the method for modeling a three-dimensional neutron physics model of a non-uniform core as described in any one of claims 1 to 5.

10. An electronic device, characterized in that: It comprises a memory and a processor, wherein the memory stores a computer program, and the processor executes the steps of the modeling method of the three-dimensional neutron physics model of the non-uniform core as claimed in any one of claims 1 to 5 by calling the computer program stored in the memory.

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