Advanced twist nuclear fuel assembly refined three-dimensional burnup analysis method and system

Abstract

The application discloses a kind of advanced twist type nuclear fuel assembly refinement three-dimensional burnup analysis method and system, the method is first obtained the physical and geometric model of advanced twist type nuclear fuel assembly, according to the geometric characteristics of advanced twist type nuclear fuel division hexahedron solid burnup area, based on multiple burnup area establishes advanced twist type nuclear fuel DAGMC geometry file, while establishing the non-structured grid file statistical result of three-dimensional advanced twist type nuclear fuel assembly, based on physical model, geometry file, non-structured grid file, burnup calculation setting spreads burnup calculation, obtains irregular twist geometry or the fuel assembly of containing burnable poison in different burnup step refinement three-dimensional space variable distribution, studies irregular geometric shape or burnable poison to the influence of assembly burnup, effectively solve the difficulty of modeling and statistical analysis in advanced twist type nuclear fuel assembly three-dimensional burnup calculation, reduce the uncertainty of calculation, provide basis for new type small reactor physical design and safety design.

Description

TECHNICAL FIELD

[0001] The application belongs to the technical field of advanced nuclear fuel of nuclear reactors, and relates to a fine three-dimensional burnup analysis method and system for advanced twisted nuclear fuel assemblies. BACKGROUND

[0002] The proportion of nuclear energy in the energy production structure directly affects the world climate, and advanced nuclear reactor fuel technology is conducive to meeting the world's energy demand and achieving the "net zero" climate goal. Fuel elements are key components of reactor cores that undergo nuclear fission and release energy. With the continuous development of innovative advanced nuclear fuel, irregular geometry and the presence of burnable poisons have higher requirements for neutron physics calculations. The fuel elements in the advanced twisted fuel assembly are twisted bodies formed by three or four petal cross sections, which have a control reactivity displacer, a uranium-zirconium alloy fuel with high thermal conductivity that can reduce fuel rod swelling, and a corrosion-resistant zirconium-niobium alloy cladding. The fuel design has inherent safety and improves economy. This design not only helps balance energy, but also addresses the issue of nuclear weapons after fuel burning and the problem of waste management during waste storage. Compared with traditional fuel, the advanced fuel has the advantages of better heat transfer performance of metal materials, mutual positioning of fuel assemblies achieved by axial helical twisting, effective promotion of coolant cross-mixing flow, improved natural circulation capacity, increased power density output, reduced heat during shutdown, improved safety margin, and the concave shape design accommodating greater irradiation swelling and reducing stress.

[0003] Accurate burnup calculation is an important part of nuclear reactor physics simulation to predict the composition of nuclear fuel and is the basis of neutron physics calculation, which has a great impact on the neutron physical properties, thermal-hydraulic characteristics, and safety analysis of the reactor core. The Monte Carlo method generally uses discrete regions as burnup regions, and each region has uniform material properties. The calculation cannot accurately predict the spatial variable distribution and its change with burnup step, and the selection of power peak factor equivalent safety margin needs to be larger, thereby limiting the performance of the reactor. In addition, for advanced twisted nuclear fuel assemblies, traditional solid geometry cannot represent the three-dimensional irregular geometric characteristics and twisting characteristics, and new fuel materials are different from conventional materials. Different absorption cross sections of dopant nuclei result in atypical spatial self-shielding effect, and the neutron flux density distribution gradient is large. The traditional modeling and burnup region division method cannot solve the problem of advanced twisted fuel burnup calculation. SUMMARY

[0004] The application aims to solve the problem of lack of three-dimensional fine burnup distribution calculation for advanced twisted nuclear fuel assemblies due to insufficient research on the neutron physical properties of advanced twisted nuclear fuel assemblies in the prior art, and provides a fine three-dimensional burnup analysis method and system for advanced twisted nuclear fuel assemblies.

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

[0006] This invention proposes an advanced method for refined three-dimensional burnup analysis of torsion-type nuclear fuel assemblies, comprising the following steps:

[0007] Obtain the physical and geometric models of the advanced torsional nuclear fuel assembly. Based on the physical and geometric models, perform geometric modeling on one nuclear fuel element in the advanced torsional nuclear fuel assembly and then divide the hexahedral solid burnup region. Convert the hexahedral solid burnup region into a DAGMC geometric file and establish the geometric model of the advanced torsional fuel assembly with the hexahedral solid burnup region based on the DAGMC file.

[0008] Obtain the burnup calculation parameters of the advanced torsional nuclear fuel assembly, establish a three-dimensional unstructured mesh result statistical file, and perform burnup calculation of the advanced torsional nuclear fuel assembly based on the physical model, the geometric model of the advanced torsional fuel assembly, the three-dimensional unstructured mesh result statistical file, and the burnup calculation parameters. Obtain the effective increment coefficient and three-dimensional refined variable distribution of each burnup step.

[0009] The effective increment coefficient and three-dimensional refined variables of each burnup step are processed to obtain the change curve of the effective increment coefficient during the lifetime, the three-dimensional refined neutron flux density distribution, fission rate distribution and nuclide density change of different energy groups in the burnup region of each burnup step, so as to realize the refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies.

[0010] Preferably, the method for dividing the hexahedral solid fuel consumption region is as follows:

[0011] A two-dimensional graphic of a fuel element in an advanced torsional nuclear fuel assembly is obtained. Based on the geometric characteristics of the two-dimensional graphic, it is divided into m quadrilateral grids. The m quadrilateral grids are then converted into m independent quadrilaterals. The quadrilaterals are swept and twisted to form m independent hexahedral solid burnup regions, thus realizing the division of the hexahedral solid burnup regions of a nuclear fuel element in a three-dimensional advanced torsional nuclear fuel assembly.

[0012] Preferably, the method for obtaining the DAGMC geometry file of the hexahedral solid burn-out region is as follows:

[0013] Based on the established geometry of an advanced torsional nuclear fuel element containing m hexahedral solid burnup regions, the coordinates and positional relationships of each hexahedron are read, the vertices of the hexahedrons are sorted, each burnup region is assigned a different material number, and the corresponding volume is calculated. The geometry of a single fuel element containing m hexahedral solid burnup regions with different volumes and different material information is converted into a DAGMC geometry file required for burnup calculation.

[0014] Preferably, the geometric model of the advanced torsional fuel assembly is as follows:

[0015] The fuel element, comprising m hexahedral solid combustion regions of different volumes and materials, is treated as a whole. In physical calculations, this forms an advanced torsional fuel assembly geometric model with an n×n arrangement. 2 ×m hexahedral burnup zones with different volumes and material information.

[0016] Preferably, the calculation parameters for the burnup of advanced torsional nuclear fuel assemblies include: neutron source distribution, particle number, total number of calculation batches, number of inactive generations, number of calculation generations per batch, number of particles per generation, calculation simulation method, burnup time step, transport-burnup coupling calculation method, and statistical variables;

[0017] The method for creating a statistical file of 3D unstructured mesh results is as follows:

[0018] The geometric arrangement of single fuel strands in the burnup regions of hexahedral entities containing m different volumes and materials is used to form an n*n advanced torsional nuclear fuel assembly three-dimensional geometry. Based on the three-dimensional geometric model, tetrahedral meshes are divided to form a three-dimensional unstructured mesh file, which is used to statistically calculate the results of three-dimensional variables.

[0019] Preferably, the method for obtaining the effective increment coefficient and three-dimensional refined variable distribution of each fuel consumption step is as follows:

[0020] Step 1: Based on the initial time or the end of the previous burnup step, obtain the three-dimensional spatial distribution of material composition and nucleon density of each hexahedral burnup region, put it into the source particle and use the probability theory method to solve the neutron transport equation, and statistically analyze the three-dimensional refined neutron flux density spatial distribution, fission reaction rate spatial distribution and reactivity multiplication coefficient of the advanced torsional fuel assembly.

[0021] Step 2, based on the n of the advanced twisted nuclear fuel assembly 2 The spatial distribution of neutron flux density and fission reaction rate in ×m burnout regions are used to solve the burnout equation and calculate the nucleon density of fuel isotopes and the nucleon density of new isotopes produced in each burnout region.

[0022] Step 3: Solve the neutron transport equation for the burnup step based on the nucleon density of the fuel isotopes in each burnup region and the nucleon density of the new isotopes produced. Repeat steps 1 and 2 to achieve the coupled calculation of the neutron transport equation and the burnup equation in sequence. After solving all burnup steps, obtain the effective increment coefficient and three-dimensional refined variable distribution for each burnup step.

[0023] Preferably, the criterion for judging the fidelity of the neutron transport equation lies in evaluating the convergence of its particle source distribution and the calculation error. A line graph of information entropy versus batch number is used to evaluate the convergence of the source distribution, and the calculation error of the effective increment factor is used to evaluate the reactivity error.

[0024] In the burnup step calculation, the effective increment coefficient of the fuel assembly at each time step is used to determine whether there is residual reactivity. If the effective increment coefficient is greater than 1, it is a supercritical design with residual reactivity that meets the design requirements. If the effective increment coefficient is less than 1, it is a subcritical state with decreasing reactivity, and the reactor cannot reach its design life.

[0025] This invention proposes an advanced three-dimensional refined burnup analysis system for torsion-type nuclear fuel assemblies, comprising:

[0026] The model building module is used to obtain the physical and geometric models of the advanced torsional nuclear fuel assembly. Based on the physical and geometric models, a nuclear fuel element in the advanced torsional nuclear fuel assembly is geometrically modeled and then a hexahedral solid burnup region is divided. The hexahedral solid burnup region is converted into a DAGMC geometric file, and a geometric model of the advanced torsional fuel assembly with the hexahedral solid burnup region is built based on the DAGMC file.

[0027] The parameter acquisition module is used to acquire the burnup calculation parameters of the advanced torsional nuclear fuel assembly, establish a three-dimensional unstructured mesh result statistical file, perform advanced torsional nuclear fuel assembly burnup calculation based on the physical model, the geometric model of the advanced torsional fuel assembly, the three-dimensional unstructured mesh result statistical file, and the burnup calculation parameters, and obtain the effective increment coefficient and three-dimensional refined variable distribution of each burnup step.

[0028] The parameter processing module is used to process the effective increment coefficient and three-dimensional refined variables of each burnup step to obtain the change curve of the effective increment coefficient during the lifetime, the three-dimensional refined neutron flux density distribution, fission rate distribution and nuclide density change of different energy groups in the burnup region of each burnup step, so as to realize the refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies.

[0029] A computer device includes a memory and a processor, the memory storing a computer program, the processor executing the computer program to implement the steps of an advanced method for refined three-dimensional burnup analysis of torsional nuclear fuel assemblies.

[0030] A computer-readable storage medium storing a computer program that, when executed by a processor, implements the steps of an advanced method for refined three-dimensional burnup analysis of torsional nuclear fuel assemblies.

[0031] Compared with the prior art, the present invention has the following beneficial effects:

[0032] This invention proposes a refined three-dimensional burnup analysis method for advanced torsional nuclear fuel assemblies. Compared to traditional regular solid geometry modeling, this method first obtains the physical and geometric models of the advanced torsional nuclear fuel assembly. Instead of using discrete regions as burnup areas, it divides the burnup region into hexahedral regions based on the geometric characteristics of the advanced fuel and establishes a geometric file. It then calculates the three-dimensional variable distribution of each burnup step with high fidelity, effectively solving the difficulties in modeling and statistical analysis in the three-dimensional neutron physics performance research of advanced torsional nuclear fuel assemblies, reducing computational uncertainty, and enabling the analysis of fuel assemblies with irregular torsional geometry or containing combustible poisons under different burnup conditions. This study refines the distribution of three-dimensional spatial variables in the burnup step, investigating the impact of irregular geometries or combustible poisons on the burnup calculation of the assembly. This includes the influence of torsional effects or torsional geometric parameters on the nuclear reactivity and neutron flux density of each burnup step, their impact on the remaining reactivity during fuel lifetime, and the neutron physics influence between mutually positioned fuel elements in advanced torsional nuclear fuel assemblies. This method effectively addresses the difficulties in statistical analysis in three-dimensional neutron physics research of advanced torsional nuclear fuel assemblies, reduces computational uncertainty, and improves the accuracy of coupled calculations involving neutron physics, thermal-hydraulic, and mechanics, providing a foundation for the physics and safety design of novel small modular reactors.

[0033] This invention proposes an advanced three-dimensional burnup analysis system for torsional nuclear fuel assemblies. By dividing the system into a model building module, a parameter acquisition module, and a parameter processing module, it achieves refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies. The modular approach ensures that each module is independent, facilitating unified management of all modules. Attached Figure Description

[0034] To more clearly illustrate the technical solutions of the embodiments of the present invention, the accompanying drawings used in the embodiments will be briefly introduced below. It should be understood that the following drawings only show some embodiments of the present invention and should not be regarded as a limitation on the scope. For those skilled in the art, other related drawings can be obtained based on these drawings without creative effort.

[0035] Figure 1 This is a flowchart of the advanced three-dimensional burnup analysis method for torsional nuclear fuel assemblies of the present invention;

[0036] Figure 2 This is a geometric model diagram of the present invention.

[0037] Figure 3 This is a diagram showing the combustion zone division of a single fuel element in this invention.

[0038] Figure 4This is a two-dimensional solid burnup region partitioning diagram of the advanced torsional nuclear fuel assembly in the open-source physics program of this invention.

[0039] Figure 5 This is a schematic diagram of the three-dimensional geometric model of the advanced torsion nuclear fuel assembly data statistics of the present invention.

[0040] Figure 6 This is a schematic diagram of the tetrahedral unstructured mesh and a partially enlarged view of the data statistics of the advanced torsional nuclear fuel assembly of the present invention.

[0041] Figure 7 This is a diagram of the advanced torsional nuclear fuel assembly refined three-dimensional burnup analysis system of the present invention. Detailed Implementation

[0042] 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.

[0043] 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.

[0044] 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.

[0045] 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.

[0046] 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.

[0047] 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.

[0048] The present invention will now be described in further detail with reference to the accompanying drawings:

[0049] This invention proposes an advanced method for refined three-dimensional burnup analysis of torsion-type nuclear fuel assemblies, such as... Figure 1 As shown, it includes the following steps:

[0050] S1. Obtain the physical and geometric models of the advanced torsional nuclear fuel assembly. Based on the physical and geometric models, perform geometric modeling on one nuclear fuel element in the advanced torsional nuclear fuel assembly and then divide the hexahedral solid burnup region. Convert the hexahedral solid burnup region into a DAGMC geometric file and establish the geometric model of the advanced torsional fuel assembly based on the DAGMC file.

[0051] The method for dividing the fuel consumption region of a hexahedral solid is as follows:

[0052] A two-dimensional graphic of a fuel element in an advanced torsional nuclear fuel assembly is obtained. Based on the geometric characteristics of the two-dimensional graphic, it is divided into m quadrilateral grids. The m quadrilateral grids are then converted into m independent quadrilaterals. The quadrilaterals are swept and twisted to form m independent hexahedral solid burnup regions, thus realizing the division of the hexahedral solid burnup regions of a nuclear fuel element in a three-dimensional advanced torsional nuclear fuel assembly.

[0053] The method for obtaining the DAGMC geometry file of the hexahedral solid burnout region is as follows:

[0054] Based on the established geometry of an advanced torsional nuclear fuel element containing m hexahedral solid burnup regions, the coordinates and positional relationships of each hexahedron are read, the vertices of the hexahedrons are sorted, each burnup region is assigned a different material number, and the corresponding volume is calculated. The geometry of a single fuel element containing m hexahedral solid burnup regions with different volumes and different material information is converted into a DAGMC geometry file required for burnup calculation.

[0055] The geometric model of the advanced torsion fuel assembly is as follows:

[0056] The fuel element, comprising m hexahedral solid combustion regions of different volumes and materials, is treated as a whole. In physical calculations, this forms an advanced torsional fuel assembly geometric model with an n×n arrangement. 2 ×m hexahedral burnup zones with different volumes and material information.

[0057] S2. Obtain the burnup calculation parameters of the advanced torsional nuclear fuel assembly, establish a three-dimensional unstructured mesh result statistical file, and perform advanced torsional nuclear fuel assembly burnup calculation based on the physical model, the geometric model of the advanced torsional fuel assembly, the three-dimensional unstructured mesh result statistical file, and the burnup calculation parameters. Obtain the effective increment coefficient and three-dimensional refined variable distribution of each burnup step.

[0058] Advanced torsion nuclear fuel assembly burnup calculation parameters include n 2 The geometric model of ×m burnup region entities, the established unstructured mesh file, neutron source distribution, particle number, total number of calculation batches, inactive generation, calculation generation per batch, number of particles per generation, power, boundary conditions, calculation simulation method, burnup chain, burnup time step, transport-burnup coupling calculation method and statistical variables;

[0059] The calculation of burnup for advanced torsional nuclear fuel assemblies requires a physical model, a geometric model of the advanced torsional fuel assembly based on a DAGMC file, a statistical file of the 3D unstructured mesh results, and burnup calculation settings. The physical model includes material parameters, power, boundary conditions, burnup chains, etc. The geometric model of the advanced torsional fuel assembly based on the DAGMC file is for establishing a 3D unstructured mesh result statistical file, and burnup calculation settings. 2 The geometric model of the hexahedral burnup region with different materials and volumes (×m), and the statistical model of the three-dimensional unstructured mesh results are the established DAGMC geometric file.

[0060] The method for creating a statistical file of 3D unstructured mesh results is as follows:

[0061] The geometric arrangement of single fuel strands in the burnup regions of hexahedral entities containing m different volumes and materials is used to form an n*n advanced torsional nuclear fuel assembly three-dimensional geometry. Based on the three-dimensional geometric model, tetrahedral meshes are divided to form a three-dimensional unstructured mesh file, which is used to statistically calculate the results of three-dimensional variables.

[0062] The method for obtaining the effective increment coefficient and three-dimensional refined variables for each fuel consumption step is as follows:

[0063] Step 1: Based on the initial time or the end of the previous burnup step, obtain the three-dimensional spatial distribution of material composition and nucleon density of each hexahedral burnup region, put it into the source particle and use the probability theory method to solve the neutron transport equation, and statistically analyze the three-dimensional refined neutron flux density spatial distribution, fission reaction rate spatial distribution and reactivity multiplication coefficient of the advanced torsional fuel assembly.

[0064] Step 2, based on the n of the advanced twisted nuclear fuel assembly 2 The spatial distribution of neutron flux density and fission reaction rate in ×m burnout regions are used to solve the burnout equation and calculate the nucleon density of fuel isotopes and the nucleon density of new isotopes produced in each burnout region.

[0065] Step 3: Solve the neutron transport equation for the burnup step based on the nucleon density of the fuel isotopes in each burnup region and the nucleon density of the new isotopes produced. Repeat steps 1 and 2 to achieve the coupled calculation of the neutron transport equation and the burnup equation in sequence. After solving all burnup steps, obtain the effective increment coefficient and three-dimensional refined variable distribution for each burnup step.

[0066] Among them, the standard for judging the fidelity of the neutron transport equation lies in evaluating the convergence of its particle source distribution and the calculation error. The convergence of the source distribution is evaluated by using a line graph of information entropy and batch number, and the calculation error of the effective increment factor is used to evaluate the reactivity error.

[0067] In the burnup step calculation, the effective increment coefficient of the fuel assembly at each time step is used to determine whether there is residual reactivity. If the effective increment coefficient is greater than 1, it is a supercritical design with residual reactivity that meets the design requirements. If the effective increment coefficient is less than 1, it is a subcritical state with decreasing reactivity, and the reactor cannot reach its design life.

[0068] S3. Process the effective increment coefficient and three-dimensional refined variables of each burnup step to obtain the change curve of the effective increment coefficient during the lifetime, the three-dimensional refined neutron flux density distribution, fission rate distribution and nuclide density change of different energy groups in the burnup region of each burnup step, and realize the refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies.

[0069] The following are the detailed steps of the advanced torsional nuclear fuel assembly refined three-dimensional burnup analysis method proposed in this invention:

[0070] Step 1: Obtain the preliminary design scheme of the advanced torsional nuclear fuel assembly, including physical and geometric models. The physical model includes parameters such as materials, boundary conditions, and temperature, while the geometric model includes parameters such as torque, length, and diameter of the advanced torsional nuclear fuel. (The geometric model is shown in the image.) Figure 2As shown, the refined three-dimensional burnup analysis method for advanced torsional nuclear fuel assemblies is applicable to geometries such as four-petaled spiral cross fuel, including but not limited to 3×3 spiral cross fuel assemblies, which can be advanced torsional nuclear fuel assemblies with arbitrary arrangement, quantity, and geometric shape.

[0071] Step 2: Preliminary work for burnup calculation of the advanced torsional nuclear fuel assembly scheme designed in Step 1, including establishing a geometric model, dividing the burnup region and assigning material information and calculation volume, and establishing a statistical file of results, as detailed below:

[0072] 1) Draw a two-dimensional diagram of a fuel element in an advanced torsional nuclear fuel assembly. Based on its geometric characteristics, divide the two-dimensional geometry into m quadrilateral grids. Use a Python script to convert the m quadrilateral grids into m independent quadrilaterals, and then sweep and torsion them to form m independent hexahedral burnup regions, such as... Figure 3 As shown, geometric modeling and burnup zone division of a nuclear fuel element in a three-dimensional advanced torsional nuclear fuel assembly are realized.

[0073] 2) Based on the advanced torsional nuclear fuel element geometry established in the first step, which includes m hexahedral solid burnup regions, a MATLAB script is used to read the coordinates and positional relationships of each hexahedron. A convex hull script is used to sort the vertices of the hexahedrons. Through secondary development, each burnup region is assigned a different material number and its corresponding volume is calculated. Finally, it is converted into a DAGMC geometric file containing m hexahedral solid burnup regions with different materials and volumes, which is required for burnup calculation. Its two-dimensional cross-section is shown below. Figure 4 As shown.

[0074] 3) Establish an advanced torsional fuel assembly geometric model. Based on the DAGMC geometric model of the fuel element in step 2), which contains m hexahedral solid combustion regions of different materials and volumes, an advanced torsional fuel assembly geometric model with an n×n arrangement is formed in the physical calculations, such as... Figure 5 As shown, the model contains n 2 ×m hexahedral burnup zones with different material numbers and volumes.

[0075] 4) Establish a statistical file for the calculation results. To facilitate the statistical analysis of the three-dimensional physical variable calculation results (including neutron flux density and fission rate), the geometric structure of the single fuel element from step 1 is arranged into an n×n distribution of an advanced torsional nuclear fuel assembly in three-dimensional geometry. Based on this three-dimensional geometric model, tetrahedral meshes are then created, such as... Figure 6 As shown, an unstructured mesh file is generated to collect statistical data.

[0076] Step 3: Advanced torsional nuclear fuel assembly burnup calculation setup. The calculation setup includes the physical model obtained in Step 1 and the setup completed in Step 2, which includes n2 The model of the advanced torsional fuel assembly with ×m burnup regions, the established unstructured mesh file, and the burnup calculation parameter settings, including neutron source distribution, particle number, total number of calculation batches, inactive generation, number of calculation generations per batch, number of particles per generation, calculation simulation mode, burnup chain, burnup time step, transport-burnup coupling calculation mode, and statistical variables.

[0077] Step 4: Based on the files and calculation settings from Steps 2 and 3, perform burnup calculations for advanced torsional nuclear fuel assemblies. The burnup calculation is a secondary development of the burnup equation based on the open-source probabilistic physics transport equations to unfold the transport-burnup coupled calculation. The burnup calculation for advanced torsional nuclear fuel assemblies includes neutron transport calculations and burnup calculations at each time step. The specific steps are as follows:

[0078] 1) Using the three-dimensional spatial distribution of material composition and nucleon density of each hexahedral burnup region known at the initial moment or solved from the burnup equation at the end of the previous burnup step as known quantities, a large number of source particles are introduced, and the neutron transport equation is solved using the probabilistic method. The three-dimensional refined neutron flux density, fission reaction rate and other spatial distribution and reactivity multiplication coefficient of the advanced torsional fuel assembly are statistically analyzed in the statistical file in step 2(4).

[0079] 2) The standard for judging the fidelity of the neutron transport equation lies in evaluating the convergence of its particle source distribution and the calculation error. The convergence of the source distribution is evaluated by using a line graph of information entropy and batch number. The effective increment factor is used to evaluate the reactive error. If the line graph region is stable, the source converges and the reactive error is within 50 pcm, then the statistical result is considered close to the true solution and the calculation continues. If it does not converge or the error is large, modify the particle number or calculation batch number and algebra set in step 2.

[0080] 3) The n of the advanced torsion-type nuclear fuel assembly calculated using 1) 2 The neutron flux density distribution and nuclear reaction rate of the ×m burnout regions are known quantities. The burnout equation is solved within the burnout time step set in step 2, and the nucleon density of the fuel isotopes in each burnout region and the nucleon density of the new isotopes produced are calculated.

[0081] 4) Based on the nucleon density of isotopes in each burnup region obtained from the burnup equation in 3), solve the neutron transport equation for this burnup step. Repeat 1) and solve all burnup steps in sequence. In the burnup step calculation, the effective increment coefficient of the fuel assembly at each time step is used to determine whether there is residual reactivity. If the effective increment coefficient is greater than 1, it is a supercritical design with residual reactivity that meets the design requirements. If the effective increment coefficient is less than 1, it is a subcritical state with decreasing reactivity and the reactor cannot reach its design life. If the life cycle does not meet the design requirements of the reactor, redesign the geometric and physical parameters in step 1.

[0082] Step 5: Normalize and process the effective increment coefficients and three-dimensional refined variable calculation results of each fuel consumption step calculated in Step 4 to obtain the life-cycle effective increment coefficient variation curve and the fuel consumption step n. 2 Three-dimensional refined neutron flux density distribution, fission rate distribution, and nuclide density variation of different energy groups in ×m burnup regions.

[0083] Among them, the fuel consumption area was divided by Python script, and the material and volume of each area were assigned and three-dimensional refined fuel consumption calculation was carried out based on the secondary development of the open source Monte Carlo program, which ensured the high fidelity of the calculation and coupling.

[0084] This invention proposes an advanced, refined three-dimensional burnup analysis system for torsional nuclear fuel assemblies, such as... Figure 7 As shown, it includes a model building module, a parameter acquisition module, and a parameter processing module;

[0085] The model building module is used to obtain the physical and geometric models of the advanced torsional nuclear fuel assembly. Based on the physical and geometric models, a nuclear fuel element in the advanced torsional nuclear fuel assembly is geometrically modeled and then a hexahedral solid burnup region is divided. The hexahedral solid burnup region is converted into a DAGMC geometric file, and a geometric model of the advanced torsional fuel assembly with the hexahedral solid burnup region is built based on the DAGMC file.

[0086] The parameter acquisition module is used to acquire the burnup calculation parameters of the advanced torsional nuclear fuel assembly, establish a three-dimensional unstructured mesh result statistical file, and perform advanced torsional nuclear fuel assembly burnup calculation based on the physical model, the geometric model of the advanced torsional fuel assembly, the three-dimensional unstructured mesh result statistical file, and the burnup calculation parameters, and obtain the effective increment coefficient and three-dimensional refined variable distribution of each burnup step;

[0087] The parameter processing module is used to process the effective increment coefficient and three-dimensional refined variables of each burnup step, and obtain the change curve of the effective increment coefficient during the lifetime, the three-dimensional refined neutron flux density distribution, fission rate distribution and nuclide density change of different energy groups in the burnup region of each burnup step, so as to realize the refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies.

[0088] The terminal device provided in this embodiment of the invention includes: a processor, a memory, and a computer program stored in the memory and executable on the processor. When the processor executes the computer program, it implements the steps in the various method embodiments described above. Alternatively, when the processor executes the computer program, it implements the functions of each module / unit in the various device embodiments described above.

[0089] The computer program can be divided into one or more modules / units, which are stored in the memory and executed by the processor to complete the present invention.

[0090] The terminal device may be a desktop computer, laptop, handheld computer, or cloud server, etc. The terminal device may include, but is not limited to, a processor and a memory.

[0091] The processor may be a central processing unit (CPU), or other general-purpose processors, digital signal processors (DSPs), application-specific integrated circuits (ASICs), field-programmable gate arrays (FPGAs), or other programmable logic devices, discrete gate or transistor logic devices, discrete hardware components, etc.

[0092] The memory can be used to store the computer program and / or module. The processor implements various functions of the terminal device by running or executing the computer program and / or module stored in the memory and calling the data stored in the memory.

[0093] If the modules / units integrated into the terminal device are implemented as software functional units and sold or used as independent products, they can be stored in a computer-readable storage medium. Based on this understanding, all or part of the processes in the methods of the above embodiments can also be implemented by a computer program instructing related hardware. The computer program can be stored in a computer-readable storage medium, and when executed by a processor, it can implement the steps of the various method embodiments described above. The computer program includes computer program code, which can be in the form of source code, object code, executable files, or certain intermediate forms. The computer-readable medium can include: any entity or device capable of carrying the computer program code, recording media, USB flash drives, portable hard drives, magnetic disks, optical disks, computer memory, read-only memory (ROM), random access memory (RAM), electrical carrier signals, telecommunication signals, and software distribution media, etc. It should be noted that the content included in the computer-readable medium can be appropriately added or removed according to the requirements of legislation and patent practice in the jurisdiction. For example, in some jurisdictions, according to legislation and patent practice, computer-readable media do not include electrical carrier signals and telecommunication signals.

[0094] This invention proposes a refined three-dimensional burnup analysis method for advanced torsional nuclear fuel assemblies. The method obtains a preliminary design scheme for advanced torsional nuclear fuel assemblies, including physical and geometric models. Prerequisites for burnup calculation include establishing the geometric model, dividing the hexahedral burnup region and assigning materials and calculation volumes, and creating a statistical file of results. The method then sets up the physical calculation settings for burnup of the advanced torsional nuclear fuel assembly. The burnup calculation is based on a secondary development of the burnup equation using open-source probabilistic physics to perform transport-burnup coupled calculations. The effective increment coefficients and three-dimensional refined variable calculation results of each burnup step are normalized and post-processed. This method provides high-fidelity calculation of the spatial variable distribution of fuel assemblies with irregular geometry or containing combustible poisons, solving the difficulties in neutron physics analysis of advanced torsional nuclear fuel assemblies, improving the accuracy of coupled calculations involving neutron physics, thermal-hydraulic, and mechanics, and providing a foundation for the physical design of novel small modular reactors. It can obtain the refined three-dimensional spatial variable distribution of fuel assemblies with irregular torsional geometry or containing combustible poisons at different burn-out steps, and study the influence of irregular geometry or combustible poisons on the calculation of assembly burn-out, including the influence of torsional effect or torsional geometry parameters on the nuclear reactivity and neutron flux density at each burn-out step, the influence on the remaining reactivity during the fuel lifetime, and the neutron physics influence between mutually positioned fuel elements in advanced torsional nuclear fuel assemblies.

[0095] The above are merely preferred embodiments of the present invention and are not intended to limit the present invention. Various modifications and variations can be made to the present invention by those skilled in the art. Any modifications, equivalent substitutions, improvements, etc., made within the spirit and principles of the present invention should be included within the scope of protection of the present invention.

Claims

1. A refined three-dimensional burnup analysis method for advanced torsional nuclear fuel assemblies, characterized in that, Includes the following steps: Obtain the physical and geometric models of the advanced torsional nuclear fuel assembly. Based on the physical and geometric models, perform geometric modeling on one nuclear fuel element in the advanced torsional nuclear fuel assembly and then divide the hexahedral solid burnup region. Convert the hexahedral solid burnup region into a DAGMC geometric file and establish the geometric model of the advanced torsional fuel assembly with the hexahedral solid burnup region based on the DAGMC file. Obtain the burnup calculation parameters of the advanced torsional nuclear fuel assembly, establish a three-dimensional unstructured mesh result statistical file, and perform burnup calculation of the advanced torsional nuclear fuel assembly based on the physical model, the geometric model of the advanced torsional fuel assembly, the three-dimensional unstructured mesh result statistical file, and the burnup calculation parameters. Obtain the effective increment coefficient and three-dimensional refined variable distribution of each burnup step. The effective increment coefficient and three-dimensional refined variables of each burnup step are processed to obtain the change curve of the effective increment coefficient during the lifetime, the three-dimensional refined neutron flux density distribution, fission rate distribution and nuclide density change of different energy groups in the burnup region of each burnup step, so as to realize the refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies. The calculation parameters for the burnup of advanced torsional nuclear fuel assemblies include: neutron source distribution, particle number, total number of calculation batches, number of inactive generations, number of calculation generations per batch, number of particles per generation, calculation simulation method, burnup time step, transport-burnup coupling calculation method, and statistical variables. The method for establishing a statistical file of three-dimensional unstructured mesh results is as follows: The geometric arrangement of single fuel strands in the fuel consumption regions of a hexahedral solid containing m different volumes and material information is formed into n... The three-dimensional geometry of the advanced torsional nuclear fuel assembly is obtained, and a three-dimensional unstructured mesh file is generated by dividing the three-dimensional geometric model into tetrahedral meshes, which is used to statistically calculate the results of three-dimensional variables. The method for obtaining the effective increment coefficient and three-dimensional refined variable distribution for each fuel consumption step is as follows: Step 1: Based on the initial time or the end of the previous burnup step, obtain the three-dimensional spatial distribution of material composition and nucleon density of each hexahedral burnup region, put it into the source particle and use the probability theory method to solve the neutron transport equation, and statistically analyze the three-dimensional refined neutron flux density spatial distribution, fission reaction rate spatial distribution and reactivity multiplication coefficient of the advanced torsional fuel assembly. Step 2, based on the n of the advanced twisted nuclear fuel assembly 2 The spatial distribution of neutron flux density and fission reaction rate in ×m burnout regions are used to solve the burnout equation and calculate the nucleon density of fuel isotopes and the nucleon density of new isotopes produced in each burnout region. Step 3: Solve the neutron transport equation for each burnup step based on the nucleon density of the fuel isotopes in each burnup region and the nucleon density of the newly produced isotopes. Repeat steps 1 and 2 to iteratively perform coupled calculations of the neutron transport equation and the burnup equation. After solving all burnup steps, obtain the effective increment coefficient and three-dimensional refined variable distribution for each burnup step. The method for dividing the hexahedral solid burnup region is as follows: A two-dimensional graphic of a fuel element in an advanced torsional nuclear fuel assembly is obtained. Based on the geometric characteristics of the two-dimensional graphic, it is divided into m quadrilateral grids. The m quadrilateral grids are then converted into m independent quadrilaterals. The quadrilaterals are swept and twisted to form m independent hexahedral solid burnup regions, thus realizing the division of the hexahedral solid burnup regions of a nuclear fuel element in a three-dimensional advanced torsional nuclear fuel assembly.

2. The method for refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies according to claim 1, characterized in that, The method for obtaining the DAGMC geometry file of the hexahedral solid burnout region is as follows: Based on the established geometry of an advanced torsional nuclear fuel element containing m hexahedral solid burnup regions, the coordinates and positional relationships of each hexahedron are read, the vertices of the hexahedrons are sorted, each burnup region is assigned a different material number, and the corresponding volume is calculated. The geometry of a single fuel element containing m hexahedral solid burnup regions with different volumes and different material information is converted into a DAGMC geometry file required for burnup calculation.

3. The method for refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies according to claim 1, characterized in that, The geometric model of the advanced torsion fuel assembly is as follows: The fuel element, comprising m hexahedral solid combustion regions of different volumes and materials, is treated as a whole. In physical calculations, this forms an advanced torsional fuel assembly geometric model with an n×n arrangement. 2 ×m hexahedral burnup zones with different volumes and material information.

4. The method for refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies according to claim 1, characterized in that, The standard for judging the fidelity of the neutron transport equation lies in assessing the convergence of its particle source distribution and the calculation error. The convergence of the source distribution is assessed by using a line graph of information entropy versus batch number, and the calculation error of the effective increment factor is used to assess the reactivity error. In the burnup step calculation, the effective increment coefficient of the fuel assembly at each time step is used to determine whether there is residual reactivity. If the effective increment coefficient is greater than 1, it is a supercritical design with residual reactivity that meets the design requirements. If the effective increment coefficient is less than 1, it is a subcritical state with decreasing reactivity, and the reactor cannot reach its design life.

5. An advanced three-dimensional refined burnup analysis system for torsional nuclear fuel assemblies, characterized in that, The advanced three-dimensional burnup analysis method for torsional nuclear fuel assemblies according to any one of claims 1 to 4 includes: The model building module is used to obtain the physical and geometric models of the advanced torsional nuclear fuel assembly. Based on the physical and geometric models, a nuclear fuel element in the advanced torsional nuclear fuel assembly is geometrically modeled and then a hexahedral solid burnup region is divided. The hexahedral solid burnup region is converted into a DAGMC geometric file, and a geometric model of the advanced torsional fuel assembly with the hexahedral solid burnup region is built based on the DAGMC file. The parameter acquisition module is used to acquire the burnup calculation parameters of the advanced torsional nuclear fuel assembly, establish a three-dimensional unstructured mesh result statistical file, perform advanced torsional nuclear fuel assembly burnup calculation based on the physical model, the geometric model of the advanced torsional fuel assembly, the three-dimensional unstructured mesh result statistical file, and the burnup calculation parameters, and obtain the effective increment coefficient and three-dimensional refined variable distribution of each burnup step. The parameter processing module is used to process the effective increment coefficient and three-dimensional refined variables of each burnup step to obtain the change curve of the effective increment coefficient during the lifetime, the three-dimensional refined neutron flux density distribution, fission rate distribution and nuclide density change of different energy groups in the burnup region of each burnup step, so as to realize the refined three-dimensional burnup analysis of advanced torsional nuclear fuel assemblies.

6. A computer device comprising a memory and a processor, wherein the memory stores a computer program, characterized in that, When the processor executes a computer program, it implements the steps of the advanced three-dimensional burnup analysis method for torsional nuclear fuel assemblies as described in any one of claims 1 to 4.

7. A computer-readable storage medium storing a computer program, characterized in that, When the computer program is executed by the processor, it implements the steps of the advanced three-dimensional burnup analysis method for torsional nuclear fuel assemblies as described in any one of claims 1 to 4.

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