A method and system for heterogeneous fuel geometry nested neutron transport and burnup calculation

Abstract

The application discloses a kind of special-shaped fuel geometric nested neutron transport and burnup calculation method and system, method includes: based on geometric nested model, the neutron transport calculation of special-shaped reactor is realized;Geometric nested model is obtained by constructing as follows: using the modeling of conventional assembly to construct entity geometry method, using direct acceleration geometry Monte Carlo method to model special-shaped fuel assembly, the geometry system space of direct acceleration geometry Monte Carlo geometry is embedded into the geometry system space of construct entity geometry, complete geometric nested modeling;After neutron transport calculation is completed, the burnup of geometric nested model is calculated, and the burnup calculation of special-shaped fuel reactor is realized.The application can improve the problem that Monte Carlo calculation cannot accurately and efficiently establish special-shaped fuel model using CSG method, improve the calculation efficiency using DAGMC method, and accurately and efficiently complete special-shaped fuel reactor core modeling and neutron transport calculation and burnup calculation.

Description

Technical Field

[0001] This invention belongs to the field of reactor core neutron transport calculation technology, specifically relating to a method and system for calculating neutron transport and burnup based on the geometric nesting of irregularly shaped fuels. Background Technology

[0002] When using reactor computational software to perform neutron transport calculations in a reactor, the first step is to efficiently and accurately establish a geometric model based on the actual materials and structural arrangement of the reactor. Subsequent transport simulation calculations are then performed on this basis. Traditional reactors, such as pressurized water reactors, use relatively regular and simple material structures. Accurate modeling can generally be achieved using traditional Monte Carlo software such as MCNP (Monte Carlo N-particle Transport Code). MCNP commonly employs Constructive Solid Geometry (CSG) for geometric modeling, which describes the model through Boolean operations on simple geometric shapes (such as cylinders, blocks, spheres, and cones). This method satisfies the geometric modeling requirements of traditional reactors.

[0003] However, with the increasing complexity of new reactor geometries and the diversification of fuel designs, including many irregularly shaped fuel reactors, traditional Monte Carlo neutron transport programs struggle to meet the demands for efficient and accurate geometric modeling. While CAD modeling is currently efficient, the development of CAD-to-CSG conversion tools addresses this issue. However, this method is time-consuming and error-prone for analyzing high-order spline surfaces, leading to the development of Monte Carlo transport methods based on unstructured meshes. Among these, surface unstructured mesh particle transport (i.e., patching the surface of a geometric entity without meshing its interior) offers better performance. A representative example is the Direct Accelerated Geometry Monte Carlo (DAGMC) method. Current Monte Carlo software such as MCNP, OpenMC, Geant4, and FLUKA all support DAGMC; however, DAGMC suffers from high memory consumption and low computational efficiency, and a satisfactory solution remains elusive. In summary, current calculations for neutron transport in irregularly shaped fuel reactors rely on relatively time-consuming and error-prone analysis of high-order spline surfaces, resulting in significant memory consumption and requiring further improvement in computational efficiency. Summary of the Invention

[0004] This invention provides a method and system for calculating neutron transport and burnup based on the geometric nesting of irregularly shaped fuels. The purpose is to solve the problems in current neutron transport calculations for irregularly shaped fuel reactors, such as the relatively time-consuming and error-prone analysis of high-order spline surfaces, large memory consumption of software, and the need for further improvement in computational efficiency.

[0005] To achieve the above objectives, the present invention adopts the following technical solution:

[0006] This invention provides a method for calculating neutron transport and burnup using geometrically nested irregular fuels, specifically comprising the following steps:

[0007] S1. Based on the geometric nesting model, a hybrid transport method of direct acceleration geometric Monte Carlo and constructed solid geometry is used. By comparing the shortest distance of the particle to the boundary in geometric Monte Carlo and constructed solid geometry in the geometric nesting model, the next geometric type that the particle enters in the geometric nesting model is determined, thereby realizing the neutron transport calculation of the heterogeneous reactor.

[0008] The geometric nesting model is constructed as follows: the irregular fuel assembly and the conventional assembly are distinguished; the conventional assembly is modeled using the constructive solid geometry method and written into the Monte Carlo input file; the irregular fuel assembly is modeled using the direct accelerated geometry Monte Carlo method and used as the input file for Monte Carlo transport calculation; the geometric system space of the direct accelerated geometry Monte Carlo geometry is embedded into the geometric system space of the constructive solid geometry to complete the geometric nesting model.

[0009] S2. After the neutron transport calculation is completed, the burnup of the geometric nesting model is calculated to realize the burnup calculation of the irregular fuel reactor.

[0010] In some implementations, S1 specifically includes:

[0011] The ray-fire function is used to calculate the particle's distance ToDAG1 from the boundary of the direct acceleration geometry Monte Carlo geometry along the flight direction, and the FliRayFire function is used to calculate the particle's distance ToCSG2 from the boundary of the constructed solid geometry along the flight direction. The total reaction cross section is calculated, and the particle's free flight path is judged. If the particle's flight distance is less than the minimum of ToDAG1 and ToCSG2, the particle transport ends; otherwise, the next judgment is performed to complete the tracking of the particle track.

[0012] Furthermore, in S1, if the condition is not met, the next judgment is performed, specifically including:

[0013] Based on the sizes of ToDAG1 and ToCSG2, determine which face the particle will enter. By calling a function, determine the next grid cell that ToDAG1 and ToCSG2 will enter respectively. Repeat the above operation to complete the tracking of the particle track.

[0014] In some implementations, in S1, in the hybrid transport method that uses direct accelerated geometric Monte Carlo and constructs solid geometry, single precision is used to perform the traversal of the bounding box operation in the ray tracing method, and a ray tracing kernel is used to complete the single precision calculation; double precision is used to calculate the solid intersections.

[0015] In some implementations, in S1, the modeling of conventional components includes performing Boolean operations on geometric surfaces.

[0016] In some implementations, in S1, modeling of the irregular fuel assembly includes: constructing a tetrahedral solid set using the direct accelerated geometric Monte Carlo method as the input file for Monte Carlo transport calculations.

[0017] Furthermore, in S1, the tetrahedral solid set is constructed as follows: an irregular fuel geometric model is constructed using CAD, preprocessed using solid modeling software to obtain a meshed geometric file; then the meshed geometric file is converted into h5m text format data, and material data is added to form a tetrahedral solid set.

[0018] In some implementations, in S2, when calculating the fuel consumption of the geometric nesting model, in the case of calculating the fuel consumption of one step, an external coupling calculation method using text file transmission is adopted to calculate the fuel consumption data of the geometric nesting model.

[0019] Furthermore, in S2, in the calculation of the fuel consumption of the nested geometry model, in the case of multi-step fuel consumption calculation, the transport and fuel consumption 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. After calculating the neutron transport under one fuel consumption step, the fuel consumption single group cross section is obtained through the reaction rate counter and used as the input parameter of the built-in ignition fuel consumption calculation module. Nucleon number density is calculated separately for each ignition fuel consumption region to obtain fuel consumption data. Nuclides in the fuel consumption data are selected as material data for the next step based on the reaction rate contribution and the density difference with the largest nucleon in the initial material.

[0020] This invention also provides a neutron transport and burnup calculation system based on the geometric nesting of irregularly shaped fuels, comprising a model building module, a neutron transport calculation module, and a burnup calculation module, wherein:

[0021] The model building module is used to embed the geometric system space of the directly accelerated geometry Monte Carlo geometry into the geometric system space of the constructed solid geometry, thus completing the geometric nesting modeling;

[0022] The neutron transport calculation module is used to determine the next geometry type that a particle enters in the geometric nesting model by comparing the shortest distance to the boundary of the particle in the geometric Monte Carlo geometry and the constructed solid geometry, based on the geometric nesting model and using a hybrid transport method of direct acceleration geometric Monte Carlo and constructed solid geometry. This enables neutron transport calculation for irregular reactors.

[0023] The burnup calculation module is used to calculate the burnup of the geometric nesting model after the neutron transport calculation is completed, thus realizing the burnup calculation of irregular fuel reactors.

[0024] Compared with existing technologies, the present invention provides a method and system for calculating neutron transport and burnup based on the geometric nesting of irregularly shaped fuels, which has the following advantages:

[0025] This invention discloses a method for calculating neutron transport and burnup using a nested geometry of irregularly shaped fuel. First, it distinguishes between conventional components and irregularly shaped fuel. Then, it models the conventional components using a crystalline spherical generator (CSG) and writes the model into a Monte Carlo input file. Next, it directly models the irregularly shaped fuel using a diffractive algebraic reactor (DAGMC), also using this model as input, thus completing the nesting of CSG and DAGMC geometries. Then, it employs a DAGMC acceleration technique based on repeating structures, and further improves particle transport efficiency using a mixed-precision ray tracing method. Finally, it uses a DAGMC-based transport and burnup coupling calculation method to quickly obtain a new round of material from the DAGMC geometry, facilitating subsequent transport calculations and improving computational efficiency. This invention addresses the problem that current Monte Carlo calculation software using the traditional CSG method cannot accurately and efficiently build irregularly shaped fuel models, improves the computational efficiency of the DAGMC method, and can relatively accurately and efficiently complete the core modeling and neutron transport calculations of irregularly shaped fuel reactors. Attached Figure Description

[0026] The accompanying drawings are provided to further understand the invention and constitute a part of this invention. The illustrative embodiments of the invention and their descriptions are used to explain the invention and do not constitute an improper limitation of the invention.

[0027] Figure 1 This is a flowchart illustrating a method for calculating neutron transport and burnup based on the geometric nesting of irregularly shaped fuels according to the present invention.

[0028] Figure 2 This is a schematic diagram of the hybrid geometry of DAGMC and CSG in the neutron transport and burnup calculation method of irregular fuel geometry nesting according to the present invention;

[0029] Figure 3 This is a schematic diagram of mixed-precision ray tracing in a method for calculating neutron transport and burnup based on the geometric nesting of irregularly shaped fuels according to the present invention;

[0030] Figure 4 This is a schematic diagram of the internal coupling fuel consumption calculation method based on DAGMC in the neutron transport and fuel consumption calculation method of the irregular fuel geometry nesting of the present invention;

[0031] Figure 5 This is a schematic diagram of the particle transport simulation process of DAGMC geometry and CSG geometry in the neutron transport and burnup calculation method of the irregular fuel geometry nesting of the present invention. Detailed Implementation

[0032] To make the objectives, technical solutions, and advantages of the embodiments of the present invention clearer, the technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. The components of the embodiments of the present invention described and shown in the accompanying drawings can generally be arranged and designed in various different configurations.

[0033] Therefore, the following detailed description of the embodiments of the invention provided in the accompanying drawings is not intended to limit the scope of the claimed invention, but merely to illustrate selected embodiments of the invention. All other embodiments obtained by those skilled in the art based on the embodiments of the invention without inventive effort are within the scope of protection of the invention.

[0034] It should be noted that similar labels and letters in the following figures indicate similar items. Therefore, once an item is defined in one figure, it does not need to be further defined and explained in subsequent figures.

[0035] In the description of the embodiments of the present invention, it should be noted that if terms such as "upper," "lower," "horizontal," or "inner" indicate the orientation or positional relationship based on the orientation or positional relationship shown in the accompanying drawings, or the orientation or positional relationship commonly used when the product of the invention is in use, they are only for the convenience of describing the present invention and simplifying the description, and do not indicate or imply that the device or element referred to must have a specific orientation, or be constructed and operated in a specific orientation, and therefore should not be construed as a limitation of the present invention. Furthermore, terms such as "first" and "second" are only used to distinguish descriptions and should not be construed as indicating or implying relative importance.

[0036] Furthermore, the use of the term "horizontal" does not imply that the component must be absolutely horizontal, but rather that it can be slightly tilted. For example, "horizontal" simply means that its direction is more horizontal than "vertical," and does not mean that the structure must be completely horizontal, but can be slightly tilted.

[0037] In the description of the embodiments of the present invention, it should also be noted that, unless otherwise explicitly specified and limited, the terms "set," "install," "connect," and "link" should be interpreted broadly. For example, they can refer to a fixed connection, a detachable connection, or an integral connection; they can refer to a mechanical connection or an electrical connection; they can refer to a direct connection or an indirect connection through an intermediate medium; and they can refer to the internal connection of two components. Those skilled in the art can understand the specific meaning of the above terms in the present invention according to the specific circumstances.

[0038] How to effectively improve the computational efficiency of the DAGMC method to complete the core modeling and neutron transport calculation of irregular fuel reactors with relatively accurate and efficient results.

[0039] like Figure 1 and Figure 2 As shown, the present invention provides a method for calculating neutron transport and burnup based on the geometric nesting of irregularly shaped fuels, comprising the following steps:

[0040] S1. Based on the geometric nesting model, a hybrid transport method of direct acceleration geometric Monte Carlo and constructed solid geometry is used. By comparing the shortest distance of the particle to the boundary in geometric Monte Carlo and constructed solid geometry in the geometric nesting model, the next geometric type that the particle enters in the geometric nesting model is determined, thereby realizing the neutron transport calculation of the heterogeneous reactor.

[0041] The geometric nesting model is constructed as follows: the irregular fuel assembly and the conventional assembly are distinguished; the conventional assembly is modeled using the constructive solid geometry method and written into the Monte Carlo input file; the irregular fuel assembly is modeled using the direct accelerated geometry Monte Carlo method and used as the input file for Monte Carlo transport calculation; the geometric system space of the direct accelerated geometry Monte Carlo geometry is embedded into the geometric system space of the constructive solid geometry to complete the geometric nesting model.

[0042] S2. After the neutron transport calculation is completed, the burnup of the geometric nesting model is calculated to realize the burnup calculation of the irregular fuel reactor.

[0043] This invention presents a method for calculating neutron transport and burnup using a nested geometry for irregularly shaped fuels. This method addresses the limitations of current CSG methods in accurately and efficiently building models of irregularly shaped fuels, and the low computational efficiency of DAGMC methods. Furthermore, this invention improves upon the limitations imposed on the design and analysis of irregularly shaped fuel assemblies and reactor cores. Firstly, this invention employs a bidirectional nesting of DAGMC and CSG geometries, distinguishing irregularly shaped fuel assemblies from relatively simple conventional components. Because the internal structure of irregularly shaped fuels in an irregularly shaped fuel reactor is irregular and requires continuous iterative optimization, DAGMC geometry is more suitable. Conversely, other components have relatively simpler structures with fewer variations, making CSG geometry more appropriate.

[0044] In geometric modeling, a CAD is used to construct a special-shaped fuel structure and embed it as a Universe (geometric system space) into the CSG geometric system. For a Universe containing DAG geometry, relevant DAGMC library functions are called to import the DAGMC geometry; for the remaining Universes, CSG-related functions are used to perform Boolean operations on geometric surfaces and execute relevant operations.

[0045] Then, two-way transport between CSG and DAGMC is implemented. According to the particle position, the geometric type where the particle is located and the set type of the next cell entered by the particle are determined. To ensure the accuracy of particle track tracing, the behavior of the particle needs to be recorded in both geometric systems simultaneously. Three functions in two geometries of DAGMC and CSG are called respectively, as shown in Table 1 below.

[0046] Table 1 Geometric functions

[0047]

[0048] As Figure 2 shown, the hybrid geometric algorithm determines whether the next geometry entered by the particle is DAGMC geometry or CSG by comparing the shortest distances of the particle to the boundaries in 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 judged 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 trace the particle transport. At the same time, during the particle transport process, the track information in two geometric systems of CSG and solid geometry needs to be saved in real time to ensure the accuracy of particle information.

[0049] The present invention can achieve more accurate modeling, and then adopts a DAGMC acceleration method based on a repetitive structure to address the problem of low DAGMC calculation efficiency. In a reactor, special-shaped fuels are generally arranged in the form of quadrilateral grids and hexagonal grids. When using the CSG method in a 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 accelerate the modeling and calculation speed. Therefore, the advantages of the CSG repetitive structure and the DAGMC complex geometry can be combined to implement a DAGMC acceleration technology based on a repetitive structure. When using a repetitive structure, first locate which lattice the neutron is in, and then only calculate the neutron flight distance for the surfaces in the underlying geometry within that lattice, thus avoiding calculating the distances for all the surfaces of all fuel assemblies in the entire reactor, thereby significantly improving the calculation efficiency.

[0050] As Figure 3 As shown, this invention employs a hybrid precision ray tracing method to improve the efficiency of ray tracing in DAGMC simulations of neutron transport. This method separates the hierarchical bounding box traversal of DAGMC from the calculation of entity intersection points. Since most computational operations in ray tracing are spent traversing hierarchical bounding boxes, using single precision to perform these operations significantly improves performance. The left side illustrates the hierarchical bounding box traversal process. First, it determines whether the particle intersects with the overall large bounding box. If they intersect, further judgment is made, subdividing the large bounding box into smaller bounding boxes based on the surface position. Finally, it determines which small bounding box the particle intersects with. These processes require numerous executions with relatively low precision, hence single precision ray tracing is used. Once the specific intersecting bounding boxes are determined, entity intersection point calculation is performed, which requires precise calculation and is executed with double precision. This improves the efficiency of ray tracing.

[0051] After performing neutron transport calculations, burnup must also be calculated to obtain material changes and changes in the reactor's effective growth factor, typically requiring multi-step burnup calculations. A hybrid transport method combining CSG and DAGMC with internally coupled fuel calculations offers several advantages over externally coupled methods using text files: First, the activation-related nuclide single-group reaction cross-sections are statistically analyzed online during neutron transport. Second, the energy spectrum is updated with each step of the multi-step calculation. Finally, the method is simpler and more convenient for users, improving the user experience.

[0052] like Figure 4 As shown, based on the DAGMC-based transport and burnup coupling calculation method, DAGMC geometry provides positioning for the statistics of burnup-related reaction rate counters. After Monte Carlo transport ends, the burnup-related single-group cross section in the reaction rate counter can be used as the input parameter for the built-in burnup calculation module. Nucleon number density is calculated for each burnup region. Then, based on the reaction rate of these nuclides, important nuclides are selected from the burnup data as updated material data, thus continuing the next Monte Carlo transport calculation.

[0053] The present invention will be further described in detail below through specific embodiments.

[0054] This invention relates to a neutron transport calculation method. It achieves accurate modeling of irregularly shaped fuels through bidirectional nesting of DAGMC and CSG geometries without affecting the modeling speed of conventional components. The calculation speed is improved by using a hybrid precision DAGMC acceleration method based on repeating structures. Furthermore, the use of a DAGMC-based transport and burnup coupling calculation method simplifies the operation and improves computational efficiency. The specific implementation steps, starting with modeling the irregularly shaped fuel reactor, are as follows:

[0055] Modeling is performed using a bidirectional nested approach of DAGMC and CSG to distinguish irregular fuels from other conventional components, with DAGMC and CSG methods modeled separately for each:

[0056] 1) For regular components, the commonly used CSG method is used for modeling. After modeling, the components are divided into different Universes. These Universes use CSG functions to perform Boolean operations on the geometric surfaces. This is part of the regular Monte Carlo modeling and is written into the Monte Carlo input file.

[0057] 2) For irregularly shaped fuel assemblies, the DAGMC method is used. First, the geometric model of the irregularly shaped fuel is accurately constructed using CAD software. Then, the solid modeling software Trelis in the DAGMC method is used for preprocessing. This software is a two-dimensional and three-dimensional finite element mesh generation tool based on solid models, mainly used to quickly generate finite element meshes for complex geometries. After processing, the CAD file is transformed into a meshed geometric file.

[0058] 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 solid set, which is also used as geometric input for Monte Carlo transport calculation.

[0059] Based on the above, when performing hybrid transport of DAGMC and CSG geometries, the Universe (geometric system space) containing DAGMC geometry is embedded into the CSG geometric system Universe to complete the nested geometric modeling. DAG geometric information and CSG geometric information are each stored using a separate set of geometric data. After the geometric modeling is completed, the hybrid transport method of DAGMC and CSG is used to accurately track particle tracks.

[0060] like Figure 5 As shown, for a sampled particle, the ray-fire function is first used to calculate the ToDAG1 (boundary distance) of the particle along the flight direction to the DAGMC, and the FliRayFire function is used to calculate the boundary distance ToCSG2 of the particle along the flight direction to the CSG. Then, the total reaction cross section is calculated, and the particle's free flight path 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 ends.

[0061] like Figure 5As shown, if the above conditions are not met, the next judgment is made. First, the particle will enter the face based on the size of ToDAG1 and ToCSG2. If ToDAG1 is smaller, the particle flies to the next face via ToDAG1, and the next cell to enter is determined by calling the next_vol (next cell) function. If ToCSG2 is smaller, the particle flies to the next face via ToCSG2, and the next cell to enter is determined by calling the ObtNextvol (get next cell) function. Then the above process is repeated to achieve accurate particle track tracking.

[0062] Specifically, a DAGMC-based acceleration method with a repetitive structure is employed to improve computational efficiency in the aforementioned transport calculations. Single-precision traversal of the bounding box operations in the ray tracing method is performed, using Intel's Embree (ray tracing library) ray tracing kernel for single-precision calculations; double-precision tracing is used to calculate entity intersections, employing the MOAB (data model library) built into DAGMC. This mixed-precision approach significantly improves computational efficiency.

[0063] Furthermore, when multi-step burnout calculations are required, using external coupling methods involving text file transfers would be cumbersome and inefficient. This invention addresses multi-step burnout calculations by employing a DAGMC-based transport and burnout internal coupling calculation method. First, the geometric position corresponding to the DAGMC is determined. Then, after calculating neutron transport within a burnout step, a burnout-related single-group cross-section is obtained using a reaction rate counter. This data can be used as input parameters for the built-in burnout calculation module. Nucleon number density is calculated separately for each burnout region to obtain burnout data.

[0064] like Figure 4 As shown, the burnup data is filtered. The burnup data contains many nuclides, including some with very small magnitudes and minimal contribution to the reaction rate calculation. Returning all nuclides directly would increase the amount of data returned and the computational burden. Therefore, nuclides that contribute significantly to the reaction rate are selected, and those within five orders of magnitude of the initial material's maximum nucleon density are chosen as the next round of material data. This invention enables accurate modeling of irregularly shaped fuels and allows for rapid neutron transport calculations for DAGMC and CSG geometries, exhibiting good computational efficiency and accuracy, and thus possessing practical significance.

[0065] Finally, it should be noted that the above description is merely a preferred embodiment of the present invention and is not intended to limit the present invention in any way. Anyone skilled in the art can readily implement the present invention according to the description and above. Any modifications, alterations, or variations made based on the disclosed technical content are equivalent embodiments of the present invention. Furthermore, any modifications, alterations, or variations made to the above embodiments based on the essential technology of the present invention are still within the protection scope of the present invention.

Claims

1. A method for heterogeneous fuel geometry nested neutron transport and burnup calculation, characterized in that, Specifically comprising the following steps: S1, based on the geometric nested model, using the hybrid transport method of direct acceleration geometry Monte Carlo and constructed entity geometry, by comparing the shortest distance of particles in the geometric nested model in the geometry Monte Carlo and the constructed entity geometry to the boundary, to determine the next geometry type of the particle entering the geometric nested model, to realize the neutron transport calculation of the special-shaped reactor; Wherein, the geometric nested model is obtained by the following construction: the special-shaped fuel assembly and the conventional assembly are distinguished, the conventional assembly is modeled using the constructed entity geometry method and written into the Monte Carlo input file, the special-shaped fuel assembly is modeled using the direct acceleration geometry Monte Carlo method and used as the input file of Monte Carlo transport calculation, the geometric system space of direct acceleration geometry Monte Carlo is embedded into the geometric system space of constructed entity geometry, and the geometric nested modeling is completed. Specifically comprising: Select ray-fire function to calculate the distance ToDAG1 of particle along the flight direction to the boundary of direct acceleration geometry Monte Carlo, select FliRayFire function to calculate the distance ToCSG2 of particle along the flight direction to the boundary of constructed entity geometry, calculate the total reaction cross section, and judge the free flight distance of particle, if the flight distance of particle meets the minimum value of ToDAG1 and ToCSG2, the particle transport is ended; If not, continue to the next judgment, determine which face the particle will enter according to the size of ToDAG1 and ToCSG2, determine the next bin of ToDAG1 and ToCSG2 by calling function, repeat the above operation, and complete the tracking of particle track; S2, after the neutron transport calculation is completed, the burnup of the geometric nested model is calculated, and the burnup calculation of the special-shaped fuel reactor is realized; In the calculation of the burnup of the geometric nested model, in the case of one-step burnup, the external coupling calculation method of the text file is used to calculate the burnup data of the geometric nested model; In the calculation of the burnup of the geometric nested model, in the case of multi-step burnup, the transport and burnup internal coupling calculation method based on direct acceleration geometry Monte Carlo geometry is used to determine the corresponding geometric position of direct acceleration geometry Monte Carlo, after the neutron transport of one burnup step is calculated, the burnup single group cross section is obtained through the reaction rate counter, which is used as the input parameter of the built-in point burnup calculation module, the neutron number density of each point burnup area is calculated respectively to obtain the burnup data, and the nuclides in the burnup data are screened as the material data of the next step according to the reaction rate contribution and the difference between the initial material maximum neutron density.

2. The method of heterogeneous fuel geometry nested neutron transport and burnup calculation of claim 1, wherein, In the S1, in the hybrid transport method of direct acceleration geometry Monte Carlo and constructed entity geometry, single precision is used to perform the traversal level bounding box operation in the ray tracing method, and the light ray tracking kernel is used to complete the single precision calculation; double precision is used to calculate the entity intersection.

3. The method of heterogeneous fuel geometry nested neutron transport and burnup calculation of claim 1, wherein, In the S1, the modeling of the conventional assembly comprises: performing by Boolean operation of geometric surface.

4. The method of heterogeneous fuel geometry nested neutron transport and burnup calculation of claim 1, wherein, In the S1, the modeling of the special-shaped fuel assembly comprises: forming a tetrahedral entity set by the direct acceleration geometry Monte Carlo method, as the input file of Monte Carlo transport calculation.

5. The method of claim 4, wherein the fuel geometry is a shaped fuel geometry. In the S1, the tetrahedral entity set is constructed in the following manner: a special-shaped fuel geometric model is constructed by CAD, is preprocessed by using entity modeling software, and a meshed geometric file is obtained; then the meshed geometric file is converted into h5m text format data, material data is added, and a tetrahedral entity set is formed.

6. A system underlying the method of neutron transport and burn-up calculation of geometrically nested shaped fuel according to any one of claims 1 to 5, characterized in that, The system comprises a model construction module, a neutron transport calculation module and a burnup calculation module, wherein: The model construction module is used for embedding a geometric system space of a direct acceleration geometry Monte Carlo geometry into a geometric system space of a constructed entity geometry, and completing geometric nested modeling; The neutron transport calculation module is used for using a mixed transport method of the direct acceleration geometry Monte Carlo and the constructed entity geometry based on the geometric nested model, comparing the shortest distances of particles in the geometric nested model to the boundaries in the geometry Monte Carlo geometry and the constructed entity geometry, judging the next geometry type into which the particles in the geometric nested model enter, and realizing neutron transport calculation of the special-shaped reactor; The burnup calculation module is used for calculating the burnup of the geometric nested model after the neutron transport calculation is completed, and realizing burnup calculation of the special-shaped fuel reactor.

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