Neutron transport and burnup calculation method and system for geometric nesting of special-shaped fuel

Abstract

The invention discloses a neutron transport and burnup calculation method and system for geometric nesting of special-shaped fuel, and the method comprises the steps: achieving the neutron transport calculation of a special-shaped reactor based on a geometric nesting model; the geometric nested model is constructed through the following steps that a conventional assembly is modeled through a solid geometry construction method, a special-shaped fuel assembly is modeled through a direct acceleration geometry Monka method, and a geometric system space of direct acceleration geometry Monka geometry is embedded into a geometric system space of solid geometry construction. Geometric nesting modeling is completed; and after neutron transport calculation is completed, calculating the burnup of the geometric nested model to realize the burnup calculation of the special-shaped fuel reactor. According to the method, the problem that a special-shaped fuel model cannot be accurately and efficiently established by using a CSG method in Monte Carlo calculation can be improved, the calculation efficiency of using a DAGMC method is improved, and special-shaped fuel reactor core modeling, neutron transport calculation and burnup calculation can be accurately and efficiently completed.

Description

Technical Field

[0001] The invention belongs to the technical field of reactor core neutron transport calculation, and in particular relates to a neutron transport and burnup calculation method and system for geometrically nested heterogeneous fuels. Background Art

[0002] When using reactor calculation software to perform reactor neutron transport calculations, the first step is to efficiently and accurately establish a geometric model based on the actual materials and arrangement structures used in the reactor, and then perform subsequent transport simulation calculations on this basis. Traditional reactors such as pressurized water reactors use relatively regular and simple material structures, and generally use traditional Monte Carlo software such as MCNP (Monte Carlo N-particle transport code) to achieve accurate modeling. MCNP generally uses constructive solid geometry (CSG for short) for geometric modeling, that is, it is described by Boolean operations of simple geometric bodies (such as cylinders, blocks, spheres, cones, etc.). In this way, the geometric modeling needs of traditional reactors can be met.

[0003] However, at present, with the increasing complexity of the geometry of new reactors, the fuel design in the reactor tends to be diversified, and there are many reactors with special fuels. The traditional Monte Carlo neutron transport program is difficult to meet the needs of efficient and accurate modeling in terms of geometric modeling. Therefore, the current CAD modeling is efficient and the method of developing CAD to CSG conversion tools is used to solve this problem. However, this method is very time-consuming and error-prone for the analysis of high-order spline surfaces. Therefore, the Monte Carlo transport method based on unstructured grids has emerged. Among them, the surface unstructured grid particle transport method (that is, the surface of the geometric entity is patched and the grid is not divided inside the entity) has a better effect. The representative one is the Direct Accelerated Geometry Monte Carlo (DAGMC) method. The current Monte Carlo software such as MCNP, OpenMC, Geant4, FLUKA, etc. all have DAGMC functions. However, DAGMC has the disadvantages of large memory consumption and low computational efficiency. There is no perfect solution yet. In summary, in the current neutron transport calculations for reactors with special fuel shapes, the analysis of high-order spline surfaces is relatively time-consuming and error-prone, the software consumes a lot of memory, and the computational efficiency needs to be further improved. Summary of the invention

[0004] The present invention provides a method and system for calculating neutron transport and burnup of geometrically nested special-shaped fuels, aiming to solve the problems that in the current neutron transport calculations for reactors with special-shaped fuels, the analysis of high-order spline surfaces is relatively time-consuming and error-prone, the software consumes a lot of memory, and the calculation efficiency needs to be further improved.

[0005] In order to achieve the above object, the present invention adopts the following technical scheme: The present invention provides a method for calculating neutron transport and burnup of geometrically nested heterogeneous fuels, which specifically comprises the following steps: S1. Based on the geometric nesting model, a hybrid transport method of direct acceleration geometric Monte Carlo and constructive solid geometry is used to determine the next geometry type that the particles in the geometric nesting model enter by comparing the shortest distance from the particle to the boundary in the geometric Monte Carlo and constructive solid geometry, so as to realize the neutron transport calculation of the special-shaped reactor. The geometric nested model is constructed as follows: distinguishing between special-shaped fuel assemblies and conventional assemblies, using constructive solid geometry to model conventional assemblies and write them into Monte Carlo input files, using direct accelerated geometry Monte Carlo method to model special-shaped fuel assemblies and use them as input files for Monte Carlo transport calculations, embedding the geometric system space of direct accelerated geometry Monte Carlo geometry into the geometric system space of constructive solid geometry, and completing geometric nested modeling; S2. After the neutron transport calculation is completed, the burnup of the geometric nested model is calculated to realize the burnup calculation of the reactor with special fuel.

[0006] In some embodiments, S1 specifically includes: The ray-fire function is used to calculate the boundary distance ToDAG1 from the particle along the flight direction to the direct acceleration geometry Monte geometry, and the FliRayFire function is used to calculate the boundary distance ToCSG2 from the particle along the flight direction to the constructed solid geometry. The total reaction cross section is calculated and the free flight distance of the particle is judged. If the flight distance of the particle satisfies the minimum value between ToDAG1 and ToCSG2, the particle transport ends; if not, the next judgment is performed to complete the tracking of the particle trajectory.

[0007] Furthermore, in S1, if the condition is not satisfied, the next judgment is performed, which specifically includes: The size of ToDAG1 and ToCSG2 is used to determine which surface the particle will enter. The next cell that ToDAG1 and ToCSG2 will enter is determined by calling a function. The above operation is repeated to complete the tracking of the particle trajectory.

[0008] In some embodiments, in S1, in a hybrid transport method using direct accelerated geometric mask and constructed solid geometry, single precision is used to perform the hierarchical bounding box traversal operation in the ray tracing method, and the ray tracing kernel is used to complete the single precision calculation; double precision is used to calculate the solid intersection.

[0009] In some embodiments, in S1 , modeling of conventional components includes: performing Boolean operations on geometric surfaces.

[0010] In some embodiments, in S1, modeling of the special-shaped fuel assembly includes: constructing a tetrahedral entity set by a direct accelerated geometric Monte Carlo method as an input file for Monte Carlo transport calculation.

[0011] Furthermore, in S1, the tetrahedron entity set is constructed as follows: a special-shaped fuel geometry model is constructed through CAD, and a solid modeling software is used for preprocessing to obtain a meshed geometry file; then the meshed geometry file is converted into data in the h5m text format, and the tetrahedron entity set is formed after adding material data.

[0012] In some embodiments, in S2, in calculating the fuel consumption of the geometric nested model, in the case of calculating the fuel consumption of one step, an external coupling calculation method transmitted by a text file is used to calculate and obtain the fuel consumption data of the geometric nested model.

[0013] Furthermore, in S2, in the calculation of the burnup of the geometric nested model, in the case of calculating multi-step burnup, the transport and burnup internal coupling calculation method based on the direct acceleration geometry Monte Carlo geometry is used to determine the geometric position corresponding to the direct acceleration geometry Monte Carlo geometry. After calculating the neutron transport under one burnup step, the burnup single group cross section is obtained through the reaction rate counter as the input parameter of the built-in ignition consumption calculation module. The nucleon number density is calculated for each ignition consumption area to obtain the burnup data. The nuclides in the burnup data are selected as the material data for the next step according to the reaction rate contribution and the density difference with the maximum nucleon of the initial material.

[0014] The present invention also provides a neutron transport and burnup calculation system for geometrically nested heterogeneous fuels, comprising a model building module, a neutron transport calculation module and a burnup calculation module, wherein: The model building module is used to embed the geometric system space of direct acceleration geometry Monte Carlo geometry into the geometric system space of constructive solid geometry to complete geometric nested modeling; The neutron transport calculation module is used to determine the next geometry type that particles in the geometric nested model enter by comparing the shortest distances from the particles in the geometric nested model to the boundary in the geometric Monte Carlo geometry and the constructive solid geometry based on the geometric nested model, so as to realize the neutron transport calculation of the special-shaped reactor. The burnup calculation module is used to calculate the burnup of the geometric nested model after the neutron transport calculation is completed, so as to realize the burnup calculation of the reactor with special fuel.

[0015] Compared with the prior art, the neutron transport and burnup calculation method and system of the present invention for geometrically nested heterogeneous fuels have the following beneficial effects: The present invention discloses a neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels. The method first distinguishes conventional components from heterogeneous fuels, then models conventional components through CSG and writes into Monte Carlo input files, then directly models heterogeneous fuels using DAGMC as input files, and completes the nesting of CSG and DAGMC geometries; then uses DAGMC acceleration technology based on repeated structures, and then uses mixed precision ray tracing methods to improve the efficiency of particle transport when using the DAGMC method; finally, a new round of materials of DAGMC geometry is quickly obtained by using a transport and burnup internal coupling calculation method based on DAGMC, so as to facilitate the completion of the next step of transport calculation, thereby improving the calculation efficiency. The present invention can improve the problem that the current Monte Carlo calculation software cannot accurately and efficiently establish a heterogeneous fuel model using the traditional CSG method, improve the calculation efficiency using the DAGMC method, and can relatively accurately and efficiently complete the modeling of heterogeneous fuel reactor cores and neutron transport calculations. BRIEF DESCRIPTION OF THE DRAWINGS

[0016] The drawings in the specification are used to provide further understanding of the present invention and constitute a part of the present invention. The schematic embodiments of the present invention and their descriptions are used to explain the present invention and do not constitute improper limitations on the present invention.

[0017] Figure 1 It is a flow chart of a method for calculating neutron transport and burnup of geometrically nested heterogeneous fuels according to the present invention; Figure 2 It is a schematic diagram of the DAGMC and CSG mixed geometry in a neutron transport and burnup calculation method for geometrically nested heterogeneous fuels of the present invention; Figure 3 A schematic diagram of mixed-precision ray tracing in a neutron transport and burnup calculation method for geometrically nested heterogeneous fuels of the present invention; Figure 4 It is a schematic flow chart of an internal coupling burnup calculation method based on DAGMC in a neutron transport and burnup calculation method for geometrically nested heterogeneous fuels of the present invention; Figure 5 It is a schematic diagram of the DAGMC geometry and CSG geometry mixed particle transport simulation process in a neutron transport and burnup calculation method for heterogeneous fuel geometry nesting of the present invention. DETAILED DESCRIPTION

[0018] In order to make the purpose, technical solutions and advantages of the embodiments of the present invention clearer, the technical solutions in the embodiments of the present invention will be clearly and completely described below in conjunction with the drawings in the embodiments of the present invention. Obviously, the described embodiments are part of the embodiments of the present invention, not all of the embodiments. Generally, the components of the embodiments of the present invention described and shown in the drawings here can be arranged and designed in various different configurations.

[0019] Therefore, the following detailed description of the embodiments of the present invention provided in the accompanying drawings is not intended to limit the scope of the invention claimed for protection, but merely represents selected embodiments of the present invention. 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.

[0020] It should be noted that similar reference numerals and letters denote similar items in the following drawings, and therefore, once an item is defined in one drawing, further definition and explanation thereof is not required in subsequent drawings.

[0021] In the description of the embodiments of the present invention, it should be noted that if the terms "upper", "lower", "horizontal", "inner", etc. indicate an orientation or positional relationship based on the orientation or positional relationship shown in the drawings, or the orientation or positional relationship in which the product of the invention is usually placed when in use, it is only for the convenience of describing the present invention and simplifying the description, and does not indicate or imply that the device or element referred to must have a specific orientation, be constructed and operated in a specific orientation, and therefore cannot be understood as a limitation on the present invention. In addition, the terms "first", "second", etc. are only used to distinguish the description, and cannot be understood as indicating or implying relative importance.

[0022] In addition, if the term "horizontal" appears, it does not mean that the component must be absolutely horizontal, but can be slightly tilted. For example, "horizontal" only means that its direction is more horizontal than "vertical", which does not mean that the structure must be completely horizontal, but can be slightly tilted.

[0023] In the description of the embodiments of the present invention, it is also necessary to explain that, unless otherwise clearly specified and limited, the terms "set", "install", "connect", and "connect" should be understood in a broad sense, for example, it can be a fixed connection, a detachable connection, or an integral connection; it can be a mechanical connection or an electrical connection; it can be a direct connection, or it can be indirectly connected through an intermediate medium, or it can be the internal connection of two components. For ordinary technicians in this field, the specific meanings of the above terms in the present invention can be understood according to specific circumstances.

[0024] How to effectively improve the computational efficiency of the DAGMC method and complete the core modeling and neutron transport calculation of special-shaped fuel reactors relatively accurately and efficiently.

[0025] like Figure 1 and Figure 2 As shown, the present invention provides a method for calculating neutron transport and burnup of geometrically nested heterogeneous fuels, comprising the following steps: S1. Based on the geometric nesting model, a hybrid transport method of direct acceleration geometric Monte Carlo and constructive solid geometry is used to determine the next geometry type that the particles in the geometric nesting model enter by comparing the shortest distance from the particle to the boundary in the geometric Monte Carlo and constructive solid geometry, so as to realize the neutron transport calculation of the special-shaped reactor. The geometric nested model is constructed as follows: distinguishing between special-shaped fuel assemblies and conventional assemblies, using constructive solid geometry to model conventional assemblies and write them into Monte Carlo input files, using direct accelerated geometry Monte Carlo method to model special-shaped fuel assemblies and use them as input files for Monte Carlo transport calculations, embedding the geometric system space of direct accelerated geometry Monte Carlo geometry into the geometric system space of constructive solid geometry, and completing geometric nested modeling; S2. After the neutron transport calculation is completed, the burnup of the geometric nested model is calculated to realize the burnup calculation of the reactor with special fuel.

[0026] The neutron transport and burnup calculation method of the geometric nesting of shaped fuels of the present invention improves the problem that the current CSG method cannot accurately and efficiently establish a shaped fuel model, and the DAGMC method has low calculation efficiency. On this basis, the present invention improves the problem of being limited in the design and analysis of shaped fuel assemblies and cores. The present invention firstly nests DAGMC geometry and CSG geometry in both directions, and distinguishes between shaped fuel assemblies and conventional relatively simple components. Because in a shaped fuel reactor, the interior of the shaped fuel is irregular and needs to be continuously iterated and optimized, so it is more suitable to use DAGMC construction geometry, while other components have relatively simple structures and less changes, and are more suitable to use CSG geometry.

[0027] During geometric modeling, CAD is used to build the special-shaped fuel structure and it is embedded into the CSG geometry system as a universe (geometry system space). For universes containing DAG geometry, the relevant DAGMC library functions are called to import DAGMC geometry; for other universes, CSG related functions are used to perform Boolean operations on geometric surfaces and perform related operations.

[0028] Then, the bidirectional transport of CSG and DAGMC is realized, and the geometry type of the particle is determined according to the particle position, as well as the collection type of the next cell that the particle enters. To ensure the accuracy of particle trajectory tracking, it is necessary to record the behavior of particles in two geometric systems at the same time, and three functions in DAGMC and CSG are called respectively, as shown in Table 1 below.

[0029] Table 1 Geometric Functions

[0030] As Figure 2 shown, the hybrid geometry algorithm determines whether the next geometry the particle enters is DAGMC geometry or CSG by comparing the shortest distances of the particle to the boundaries under two different geometric types. The DAGMC geometry is embedded into the CSG geometry. When the particle is at point A, its distance to the next DAGMC boundary is ToDAG1, and its distance to the next CSG boundary is ToCSG2. Since ToDAG1 < ToCSG2, it can be determined through judgment that the particle is still in the DAGMC geometry. When the particle is in different geometric systems, the program will call the corresponding functions to track the particle transport. At the same time, during the particle transport process, it is necessary to save the track information in the two geometric systems of CSG and solid geometry in real time to ensure the accuracy of the particle information.

[0031] The present invention can achieve more accurate modeling, and then adopts a DAGMC acceleration method based on a repetitive structure to make up for the problem of low DAGMC calculation efficiency. In a reactor, the shaped fuel is generally arranged in the form of quadrilateral grids and hexagonal grids. When using the CSG method in the Monte Carlo program, the grid geometry can be modeled and neutron transport can be carried out in the form of a repetitive structure (lattice), which can speed up the modeling and calculation speed. Therefore, the advantages of the CSG repetitive structure and the DAGMC complex geometry can be combined to realize the DAGMC acceleration technology based on the repetitive structure. When adopting the repetitive structure, first locate which grid the neutron is in, and then only calculate the neutron flight distance for the surfaces in the underlying geometry within that grid, thus avoiding calculating the distances for all the surfaces of all fuel assemblies in the entire reactor, thereby significantly improving the calculation efficiency.

[0032] As Figure 3 shown, the present invention adopts a method of hybrid-precision ray tracing to improve the efficiency of ray tracing when DAGMC simulates neutron transport. This method separates the traversal of the hierarchical bounding boxes of DAGMC from the calculation of the intersections with the entities. Since most of the computational operations in ray tracing are spent on traversing the hierarchical bounding boxes, using single-precision to perform these operations can greatly improve the performance. The left side is the process of traversing the hierarchical bounding boxes. First, it is judged whether the particle intersects with the overall large bounding box. If it intersects, further judgment is made. The large bounding box is subdivided into small bounding boxes according to the surface position, and finally it is determined which small bounding box the particle intersects with. These processes need to be executed many times and require relatively low precision, so single-precision is used for ray tracing. When the specific intersecting bounding box is determined, subsequent entity intersection calculations need to be performed, and this calculation requires precise calculation and is executed in double-precision. In this way, the efficiency of ray tracing is improved.

[0033] After doing the neutron transport calculation, the fuel consumption must also be calculated to obtain the changes in materials and the changes in the effective value-added factor of the reactor, and multi-step fuel consumption calculations are required. The internal coupling calculation method of transport and fuel consumption that combines the CSG and DAGMC mixed transport methods with the internal coupling fuel calculation is used. Compared with the external coupling mode that uses text files for transmission, the internal coupling calculation method of transport and fuel consumption: First, the activation-related single-group reaction cross sections of nuclides are online statistically calculated during the neutron transport process. Secondly, in multi-step calculations, the energy spectrum will be updated with each step of the calculation. Finally, the user operation is simpler and more convenient, which improves the user experience.

[0034] like Figure 4 As shown in the figure, based on the DAGMC transport and burnup internal coupling calculation method, the DAGMC geometry provides positioning for the statistics of the burnup-related reaction rate counter. After the Monte Carlo transport is completed, the burnup-related single group cross section in the reaction rate counter can be used as an input parameter for the built-in ignition consumption calculation module. The nucleon number density is calculated for each ignition consumption area, and then according to the reactivity of these nuclides, important nuclides are selected from the burnup data as updated material data, so as to continue the next Monte Carlo transport calculation.

[0035] The present invention is further described in detail below through specific examples.

[0036] The present invention is a neutron transport calculation method, which realizes accurate modeling of special-shaped fuel by bidirectional nesting of DAGMC geometry and CSG geometry, without affecting the modeling speed of conventional components. The mixed precision DAGMC acceleration method based on repeated structure is used to improve the neutron transport calculation speed. The transport and burnup internal coupling calculation method based on DAGMC is used to make the operation simpler and improve the calculation efficiency. Starting from the modeling of special-shaped fuel reactor, the specific implementation steps are as follows: Use DAGMC and CSG bidirectional nesting methods to model, distinguish special-shaped fuel and other conventional parts, and use DAGMC method and CSG method to model respectively: 1) For conventional components, the commonly used CSG method is used for modeling. After the modeling is completed, it is divided into different universes. These universes all use CSG functions to perform related operations on Boolean operations of geometric surfaces. They are part of conventional Monte Carlo modeling and written into Monte Carlo input files.

[0037] 2) For special-shaped fuel components, the DAGMC method is used. First, the special-shaped fuel geometric model is accurately constructed using CAD software, and then pre-processed using the solid modeling software Trelis in the DAGMC method. This software is a two-dimensional and three-dimensional finite element mesh generation tool based on solid models. It is mainly used to quickly generate finite element meshes of complex geometric bodies. After processing, the CAD file becomes a meshed geometry file.

[0038] 3) After obtaining the mesh geometry file, use the MOAB (file format conversion) function in DAGMC to convert the mesh file into h5m text format data; then add material and other data to construct a tetrahedral entity set, which is also used as a geometric input for Monte Carlo transport calculations.

[0039] Based on the above, when doing DAGMC and CSG geometry mixed transport, the universe (geometry system space) containing DAGMC geometry is embedded into the CSG geometry system universe to complete the geometry nesting modeling. DAG geometry information and CSG geometry information are stored separately using a set of geometry data. After the geometry modeling is completed, the DAGMC and CSG mixed transport method is used to accurately track the particle trajectory.

[0040] like Figure 5 As shown, for a sampled particle, the ray-fire function is first used to calculate the ToDAG1 (boundary distance) from the particle to DAGMC along the flight direction, and the FliRayFire function is used to calculate the boundary distance ToCSG2 from the particle to CSG along the flight direction. Then the total reaction cross section is calculated, and the particle free flight distance is sampled and judged. If the flight distance is less than the minimum value of ToDAG1 and ToCSG2, it means that the particle can reach the collision point within the boundary and the particle transport is completed.

[0041] like Figure 5 As shown, if the above conditions are not met, the next judgment is carried out. First, the size of ToDAG1 and ToCSG2 is used to determine which surface the particle will enter. If ToDAG1 is small, the particle flies ToDAG1 to the next surface, and the next cell that the particle enters is determined by calling the next_vol (next cell) function; if ToCSG2 is small, the particle flies ToCSG2 to the next surface, and the next cell that the particle enters is determined by calling the ObtNextvol (get next cell) function; then the above process is repeated to achieve accurate tracking of particle trajectories.

[0042] Specifically, the DAGMC acceleration method based on repeated structure is used to improve the calculation efficiency in the above transport calculation. Single precision is used to perform the hierarchical bounding box traversal operation in the ray tracing method, and Intel's Embree (ray tracing library) ray tracing kernel is used to complete the single precision calculation; double precision is used to calculate the entity intersection, and the MOAB (data model library) provided by DAGMC is used for calculation. The mixed precision calculation method is used to significantly improve the calculation efficiency.

[0043] In addition, when multi-step burnup is generally required, if the external coupling method of text file transmission is used, it will be cumbersome and inefficient. When facing multi-step burnup calculation, the present invention adopts the transport and burnup internal coupling calculation method based on DAGMC. First, the geometric position corresponding to DAGMC is determined, and then the neutron transport under a burnup step is calculated. The burnup-related single group cross section is obtained through the reaction rate counter. These data can be used as input parameters of the built-in ignition consumption calculation module. The nucleon number density is calculated for each ignition consumption area to obtain the burnup data.

[0044] like Figure 4 As shown, the burnup data is screened. There are many types of nuclides in the burnup data, including nuclides with very small orders of magnitude and little contribution to the reaction rate calculation. If all nuclides are returned directly, the amount of returned data and the amount of calculation will increase. Select nuclides with large contribution to the reaction rate, and select nuclides with a difference of 5 orders of magnitude from the maximum nucleon density of the initial material as the new round of material data. The present invention can realize accurate modeling of special-shaped fuels, and can realize fast DAGMC geometry and CSG geometry neutron transport calculations, and has better calculation efficiency and accuracy, and has certain practical significance.

[0045] Finally, it should be noted that the above is only a preferred embodiment of the present invention and does not limit the present invention in any form. Any ordinary technician in the industry can smoothly implement the present invention as shown in the specification and described above, and any equivalent changes, modifications and evolutions made by using the technical content disclosed above are all equivalent embodiments of the present invention. At the same time, any equivalent changes, modifications and evolutions made to the above embodiments based on the essential technology of the present invention are still within the protection scope of the technical solution of the present invention.

Claims

1. A method for calculating neutron transport and burnup of geometrically nested heterogeneous fuels, characterized in that: The specific steps include: S1. Based on the geometric nesting model, a hybrid transport method of direct acceleration geometric Monte Carlo and constructive solid geometry is used to determine the next geometry type that the particles in the geometric nesting model enter by comparing the shortest distance from the particle to the boundary in the geometric Monte Carlo and constructive solid geometry, so as to realize the neutron transport calculation of the special-shaped reactor. The geometric nested model is constructed as follows: distinguishing between special-shaped fuel assemblies and conventional assemblies, using constructive solid geometry to model conventional assemblies and write them into Monte Carlo input files, using direct accelerated geometry Monte Carlo method to model special-shaped fuel assemblies and use them as input files for Monte Carlo transport calculations, embedding the geometric system space of direct accelerated geometry Monte Carlo geometry into the geometric system space of constructive solid geometry, and completing geometric nested modeling; S2. After the neutron transport calculation is completed, the burnup of the geometric nested model is calculated to realize the burnup calculation of the reactor with special fuel.

2. The neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels according to claim 1 is characterized in that: The S1 specifically includes: The ray-fire function is used to calculate the boundary distance ToDAG1 from the particle along the flight direction to the direct acceleration geometry Monte geometry, and the FliRayFire function is used to calculate the boundary distance ToCSG2 from the particle along the flight direction to the constructed solid geometry. The total reaction cross section is calculated and the free flight distance of the particle is judged. If the flight distance of the particle satisfies the minimum value between ToDAG1 and ToCSG2, the particle transport ends; if not, the next judgment is performed to complete the tracking of the particle trajectory.

3. The neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels according to claim 2 is characterized in that: In S1, if the condition is not satisfied, the next step of the judgment specifically includes: The size of ToDAG1 and ToCSG2 is used to determine which surface the particle will enter. The next cell that ToDAG1 and ToCSG2 will enter is determined by calling a function. The above operation is repeated to complete the tracking of the particle trajectory.

4. The neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels according to claim 1 is characterized in that: In S1, in a hybrid transport method using direct accelerated geometric mask and constructed solid geometry, single precision is used to perform the hierarchical bounding box traversal operation in the ray tracing method, and a ray tracing kernel is used to complete single precision calculations; double precision is used to calculate solid intersections.

5. The neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels according to claim 1 is characterized in that: In said S1, modeling of conventional components includes: performing Boolean operations on geometric surfaces.

6. The neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels according to claim 1 is characterized in that: In the S1, the modeling of the special-shaped fuel assembly includes: constructing a tetrahedral entity set by a direct accelerated geometric Monte Carlo method as an input file for Monte Carlo transport calculation.

7. The neutron transport and burnup calculation method for geometrically nested heterogeneous fuels according to claim 6, characterized in that: In S1, the tetrahedron entity set is constructed as follows: a special-shaped fuel geometry model is constructed through CAD, and pre-processed using solid modeling software to obtain a meshed geometry file; then the meshed geometry file is converted into data in h5m text format, and material data is added to form a tetrahedron entity set.

8. The neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels according to claim 1 is characterized in that: In the above S2, in calculating the fuel consumption of the geometric nested model, in the case of calculating the fuel consumption of one step, an external coupling calculation method transmitted by a text file is used to calculate and obtain the fuel consumption data of the geometric nested model.

9. The neutron transport and burnup calculation method for geometric nesting of heterogeneous fuels according to claim 8, characterized in that: In the S2, in the calculation of the burnup of the geometric nested model, in the case of calculating multi-step burnup, the transport and burnup internal coupling calculation method based on the direct acceleration geometry Monte Carlo geometry is adopted to determine the geometric position corresponding to the direct acceleration geometry Monte Carlo geometry, and after calculating the neutron transport under one burnup step, the burnup single group cross section is obtained through the reaction rate counter as the input parameter of the built-in ignition consumption calculation module, and the nucleon number density is calculated for each ignition consumption area to obtain the burnup data, and the nuclides in the burnup data are selected as the material data for the next step according to the reaction rate contribution and the density difference with the maximum nucleon of the initial material.

10. A system based on the neutron transport and burnup calculation method for geometric nesting of special-shaped fuels as described in any one of claims 1 to 9, characterized in that: The system includes a model building module, a neutron transport calculation module and a burnup calculation module, among which: The model building module is used to embed the geometric system space of direct acceleration geometry Monte Carlo geometry into the geometric system space of constructive solid geometry to complete geometric nested modeling; The neutron transport calculation module is used to determine the next geometry type that particles in the geometric nested model enter by comparing the shortest distances from the particles in the geometric nested model to the boundary in the geometric Monte Carlo geometry and the constructive solid geometry based on the geometric nested model, so as to realize the neutron transport calculation of the special-shaped reactor. The burnup calculation module is used to calculate the burnup of the geometric nested model after the neutron transport calculation is completed, so as to realize the burnup calculation of the reactor with special fuel.

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