Fast spectrum small nuclear reactor core transient multi-physics field coupling method and computer program product

Abstract

The present invention relates to a fast-spectrum small-scale nuclear reactor core transient multi-physics field coupling method and computer program product to address the technical problems of existing nuclear reactor transient analysis, which considers a single factor, has low reliability, and cannot achieve accurate and comprehensive evaluation. The method includes: 1. Constructing a neutron microscopic cross-section database for the nuclear reactor core; constructing a neutron transport kinetics, thermal hydraulics, and mechanical calculation grid model; setting physical property parameters for thermal hydraulics and mechanical calculations; 2. Setting a preset total simulation time and determining initial and boundary conditions for the calculation; 3. Time stepping to update the current time; calculating the power density distribution, temperature distribution, density distribution of the core at the current time, and the displacement of each node of the core and reflector calculation grid, iterating until the convergence limit is met, and determining whether it is the final time. If so, the calculation ends; if not, time stepping continues until the calculation ends.

Description

Technical Field

[0001] The present invention relates to a nuclear reactor transient numerical analysis method, and in particular to a fast spectrum small nuclear reactor core transient multi-physical field coupling method and a computer program product. Background Art

[0002] Fast-spectrum small nuclear reactors have technical advantages such as high energy density, high mobility, fast deployment speed and long life. They not only have great potential in solving major problems such as the world's growing energy demand and global warming, but are also the best choice for special energy in future areas such as emergency power supply on land, deep sea, and deep space. They can fundamentally solve the power shortcomings of existing equipment.

[0003] Due to the compact design of fast-spectrum small reactors (FSSRs), dense fuel distribution, small core size, and long mean free path of fast neutrons in the core, core neutron leakage is strong, and the reactivity feedback from the Doppler effect of the core fuel is small. The main source of fuel reactivity feedback is geometric deformation caused by mechanical effects under temperature or external loads. Furthermore, due to the strong core neutron leakage, the reactivity feedback coefficient of the geometric deformation of the core reflector is also large. The temperature rise of the reflector is different from that of the core fuel, with a time delay. Affected by the core heat transfer process, accurate simulation of the reflector has a significant impact on the accuracy of transient calculations. The application environment is complex. For example, land-based mobile fast-spectrum small reactors are more likely to be affected by mechanical effects such as impact and vibration than traditional reactors. These external loads also affect the core geometry and its neutronic properties. Therefore, considering the mechanical effects under temperature and external loads, conducting core transient characteristics analysis based on the multi-physics coupling of neutronics, thermal hydraulics, and mechanics is of great significance to the operational safety of fast-spectrum small nuclear reactors.

[0004] Currently, transient characteristics analysis of fast-spectrum small nuclear reactor cores generally approximates deformation caused by mechanical effects as axial and radial deformation, failing to account for localized non-uniform deformation. The impact of mechanical deformation on core reactivity generally equates geometric changes to changes in nucleon density, posing the risk of non-strict equivalence during transient calculations. Considering temperature loads alone does not meet the requirements for transient safety characteristics analysis in complex fast-spectrum small nuclear reactor applications. In summary, existing multi-physics calculation methods for transient fast-spectrum small nuclear reactor cores lack sufficient credibility, and a multi-physics coupling method that can accurately and comprehensively evaluate the transient safety characteristics of fast-spectrum small nuclear reactor cores is still lacking. Summary of the Invention

[0005] In order to solve the technical problems that the existing fast spectrum transient analysis of small nuclear reactors considers only a single factor, the credibility of the results is low, and it is impossible to accurately and comprehensively evaluate the transient safety characteristics of the nuclear reactor core, the present invention proposes a fast spectrum small nuclear reactor core transient multi-physics field coupling method and computer program product.

[0006] The technical solution provided by the present invention is:

[0007] A fast spectrum small nuclear reactor core transient multi-physics field coupling method is special in that it includes the following steps:

[0008] S1. Construct a neutron microscopic cross-section database for nuclear reactor cores;

[0009] At the same time, a neutron transport mechanics calculation grid model, a thermal hydraulic calculation grid model, and a mechanical calculation grid model are constructed; the neutron transport mechanics calculation grid model is defined as the initial neutron transport mechanics calculation grid, the thermal hydraulic calculation grid model is defined as the initial thermal hydraulic calculation grid, and the mechanical calculation grid model is defined as the initial mechanical calculation grid;

[0010] At the same time, the physical parameters of thermal hydraulic calculation and mechanical calculation are set based on the material of the reactor core;

[0011] S2. Setting a preset total simulation time and determining initial conditions and boundary conditions for calculation based on the reactor's initial time operating state; the initial conditions include initial temperature distribution, initial density distribution, and initial node displacement distribution;

[0012] S3. Time stepping to update the current time; based on the neutron microscopic cross-section database, the neutron transport dynamics calculation grid model, the thermal-hydraulic calculation grid model, the mechanical calculation grid model, the physical parameters of the thermal-hydraulic calculation and the physical parameters of the mechanical calculation in step S1, the power density distribution, temperature distribution, density distribution of the core and the displacement of each node of the core and reflector calculation grid at the current time are calculated, and iteration is performed until the power density distribution, temperature distribution, density distribution and the displacement of each node of the core and reflector calculation grid meet the set convergence limit, then it is determined whether the final time of the preset total simulation time has been reached. If the set final time has been reached, the calculation is terminated; if the final time has not been reached, time stepping is continued to update the current time until the set final time is reached, the calculation is terminated, and the transient characteristics of the nuclear reactor core within the preset total simulation time are obtained.

[0013] Furthermore, step S3 is specifically as follows:

[0014] S3.1. Obtain historical temperature distribution, where the historical temperature is the initial temperature distribution; obtain historical density distribution, where the historical density distribution is the initial density distribution; obtain historical grids, where the historical grids are the initial neutron transport kinetics calculation grids and the initial thermal hydraulic calculation grids; obtain historical node displacement distributions, where the historical node displacement distributions are the initial node displacement distributions;

[0015] S3.2. Execute the first loop system until the first end condition is met;

[0016] The first circulation system comprises:

[0017] A1. Time stepping to obtain the current time;

[0018] A2. Execute the second loop system until the second end condition is met;

[0019] A3. Determine whether to execute the next time step;

[0020] The first end condition is: determining whether the current time is the end time of the preset total simulation time, if so, the calculation ends; if not, returning to step A1 until the calculation ends and the transient characteristics of the nuclear reactor core within the preset total simulation time are obtained;

[0021] The second circulation system comprises:

[0022] B1. Based on the neutron microscopic cross-section database of step S1, according to the historical temperature distribution and historical density distribution of the fuel and reflector layer, a bilinear interpolation method is used to calculate and obtain the current microscopic reaction cross section of each nuclide;

[0023] According to the historical density distribution and the historical node displacement distribution, the current nucleon density is obtained;

[0024] Calculate the neutron macroscopic reaction cross section of the fuel and reflector corresponding materials based on the current microscopic reaction cross section and current nucleon density of each nuclide;

[0025] B2. Perform neutron transport kinetics calculations based on the neutron macroscopic reaction cross sections of the fuel and reflector materials obtained in step B1 and the corresponding neutron transport kinetics calculation grid in the history grid to obtain the core power density distribution;

[0026] B3. Using the interpolation method, calculate the heat source term Q required for the corresponding thermal hydraulic calculation based on the core power density distribution obtained in step B2;

[0027] B4. Based on the core power density distribution obtained in step B2, the heat source term Q obtained in step B3, and the corresponding thermal-hydraulic calculation grid in the historical grid, and based on the physical property parameters calculated by the thermal-hydraulic calculation, thermal-hydraulic calculations are performed on the fuel and reflector layer in the core to obtain the current temperature distribution, current density distribution, and current pressure distribution of the core; the current temperature distribution and current density distribution are respectively updated as the historical temperature distribution and historical density distribution of the next iteration;

[0028] B5. Based on the current temperature distribution, current pressure distribution, current external force load, and initial mechanical calculation grid of the core in step B4, and based on the physical property parameters obtained from the mechanical calculation, mechanical calculations are performed on the core and reflector to obtain the current node displacements of the core and reflector on the calculation grid. These displacements are updated as the historical node displacements for the next iteration, and the historical grid for the next iteration is obtained using the updated historical node displacements.

[0029] The second termination condition is: judging whether the power density distribution, temperature distribution, density distribution and node displacement of the core have converged at the current time; if the set convergence limit is not met, returning to step B1 until the set convergence limit is met; if the set convergence limit is met, executing step A3.

[0030] Furthermore, step S1 is specifically as follows:

[0031] A Monte Carlo program is used to construct a neutron micro-cross-section database consisting of corresponding neutron micro-cross-sections at different state points based on the state parameters of fuel temperature, fuel density, coolant temperature, coolant density, reflector temperature and reflector density at different state points through the evaluation database;

[0032] According to the geometry and material of the reactor core, the neutron transport kinetics calculation grid model, the thermal hydraulic calculation grid model and the mechanical calculation grid model are constructed based on the meshing software.

[0033] Furthermore, in step B1, the solution formula for the neutron macroscopic reaction cross section is:

[0034]

[0035] Where Σ is the neutron macroscopic cross section, N k is the current nuclear density of the kth nuclide; σ k is the current microscopic reaction cross section of the kth nuclide.

[0036] Furthermore, in step B2, the neutron transport kinetics calculation formula is:

[0037]

[0038] in, is the neutron angular flux at position r, with energy E, moving direction Ω, and time t; ∑(r,E) is the neutron macroscopic reaction cross section at position r, with energy E; q s (r,Ω,E,t) is the scattered neutron source at position r, with energy E, moving direction Ω and time t; q f (r,Ω,E,t) is the fission neutron source at position r, energy E, direction Ω, and time t.

[0039] Furthermore, in step B4, the thermal hydraulic calculation formula is:

[0040]

[0041]

[0042]

[0043]

[0044] Among them, ρ L is the fluid density, u L is the fluid velocity, p is the pressure, τ is the viscous stress tensor, f L is the generalized momentum source term, h is the fluid specific enthalpy, k L is the thermal conductivity of the fluid, T L is the fluid temperature, T s is the fluid temperature, ρ s is the solid density, c p is the specific heat capacity, ρ s is the solid density, k s is the solid thermal conductivity, Q is the heat source term;

[0045] In step B5, the mechanical calculation formula for the core and reflector is:

[0046]

[0047]

[0048] σ=D:ε

[0049] u s is the node displacement, σ is the stress tensor, f s is the external load, Tr represents the transpose, D is the elastic matrix; ε is the strain tensor.

[0050] Furthermore, the convergence judgment formula is specifically:

[0051]

[0052]

[0053]

[0054]

[0055]

[0056]

[0057]

[0058]

[0059] in, are the maximum errors of power density, liquid temperature, solid temperature, liquid density, solid density, x-direction node displacement, y-direction node displacement, and z-direction node displacement; v Q 、 are the maximum allowable errors of power density, liquid temperature, solid temperature, liquid density, solid density, and nodal displacement, respectively, that is, the convergence limits corresponding to power density, liquid temperature, solid temperature, liquid density, solid density, and nodal displacement, respectively; n+1 is the n+1th time step, i is the i-th iteration, and i+1 is the i+1th iteration.

[0060] Furthermore, in step S1, the Monte Carlo program is OpenMC or MCNP;

[0061] The evaluation database is the China Evaluation Nuclear Database CENDL or the United States Evaluation Nuclear Database ENDF;

[0062] The mesh generation software is Gmesh, ICEM or Pointwise.

[0063] The present invention also provides a computer program product having a computer program stored thereon, characterized in that when the computer program is executed by a processor, the steps of the above-mentioned fast spectrum small nuclear reactor core transient multi-physical field coupling method are implemented.

[0064] Beneficial effects of the present invention:

[0065] 1. The fast-spectrum small nuclear reactor core transient multi-physics field coupling method provided by the present invention simulates the non-uniform deformation of the core through the coupled calculation of neutron transport kinetics, thermal hydraulics and mechanics in the transient process, thereby realizing the evaluation and analysis of the impact of material changes and non-uniform deformation on the reactivity of the fast-spectrum small nuclear reactor core in the transient process; at the same time, the transient effects of temperature load (temperature distribution) and external force load (pressure distribution) are fully considered in the coupled calculation, so that the transient safety characteristics evaluation of the nuclear reactor core is accurate and comprehensive.

[0066] 2. The fast spectrum small nuclear reactor core transient multi-physics field coupling method provided by the present invention comprehensively considers the influence of the mechanical deformation effects of the fuel and reflector layer in the core on the transient characteristics of the core.

[0067] 3. The neutron transport dynamics calculation grid model, thermal hydraulic calculation grid model and mechanical calculation grid model in the fast spectrum small nuclear reactor core transient multi-physics field coupling method provided by the present invention are all based on unstructured grid modeling, which does not limit the geometric shape of the nuclear reactor core and has a wide range of applications. BRIEF DESCRIPTION OF THE DRAWINGS

[0068] Figure 1 This is a flow chart of an embodiment of the fast spectrum small nuclear reactor core transient multi-physics field coupling method of the present invention. DETAILED DESCRIPTION

[0069] See also Figure 1 This embodiment provides a fast spectrum small nuclear reactor core transient multi-physics field coupling method, characterized by comprising the following steps:

[0070] S1. Using a Monte Carlo program, according to the state parameters of fuel temperature, fuel density, coolant temperature, coolant density, reflector temperature and reflector density at different state points, a neutron micro-cross-section database consisting of corresponding neutron micro-cross-sections at different state points is constructed through an evaluation database; the Monte Carlo program includes but is not limited to OpenMC or MCNP, and the evaluation database includes but is not limited to the China Nuclear Evaluation Database CENDL or the United States Nuclear Evaluation Database ENDF.

[0071] At the same time, according to the geometry and material of the reactor core, a neutron transport dynamics calculation grid model, a thermal hydraulic calculation grid model, and a mechanical calculation grid model are constructed based on grid division software, wherein the grid division software includes but is not limited to Gmesh, ICEM, or Pointwise; the neutron transport dynamics calculation grid model is defined as an initial neutron transport dynamics calculation grid, the thermal hydraulic calculation grid model is defined as an initial thermal hydraulic calculation grid, and the mechanical calculation grid model is defined as an initial mechanical calculation grid;

[0072] At the same time, the physical parameters for thermal hydraulic calculations and the physical parameters for mechanical calculations are set based on the materials of the reactor core.

[0073] S2. Setting a preset total simulation time and determining initial conditions and boundary conditions for calculation based on the reactor's initial time operating state; the initial conditions include initial temperature distribution, initial density distribution, and initial node displacement distribution;

[0074] S3, time stepping, updating the current time;

[0075] Based on the neutron microscopic cross-section database, the neutron transport dynamics calculation grid model, the thermal-hydraulic calculation grid model, the mechanical calculation grid model, the physical property parameters of the thermal-hydraulic calculation, and the physical property parameters of the mechanical calculation in step S1, the power density distribution, temperature distribution, density distribution of the core, and the displacement of each node of the core and reflector calculation grid at the current time are calculated, and iteration is performed until the power density distribution, temperature distribution, density distribution, and the displacement of each node of the core and reflector calculation grid meet the set convergence limit, then it is determined whether the final time of the preset total simulation time has been reached; if the set final time has been reached, the calculation is terminated; if the final time has not been reached, time stepping is continued, the current time is updated, until the set final time is reached, the calculation is terminated, and the transient characteristics of the nuclear reactor core within the preset total simulation time are obtained.

[0076] Specifically, S3.1, obtaining a historical temperature distribution, where the historical temperature is the initial temperature distribution; obtaining a historical density distribution, where the historical density distribution is the initial density distribution; obtaining a historical grid, where the historical grid is the initial neutron transport kinetics calculation grid and the initial thermal hydraulic calculation grid; obtaining a historical node displacement distribution, where the historical node displacement distribution is the initial node displacement distribution;

[0077] S3.2. Execute the first loop system until the first end condition is met;

[0078] The first circulation system comprises:

[0079] A1. Time stepping to obtain the current time;

[0080] A2. Execute the second loop system until the second end condition is met;

[0081] A3. Determine whether to execute the next time step;

[0082] The first end condition is: determining whether the current time is the end time of the preset total simulation time, if so, the calculation ends; if not, returning to step A1 until the calculation ends and the transient characteristics of the nuclear reactor core within the preset total simulation time are obtained;

[0083] The second circulation system comprises:

[0084] B1. Based on the neutron microscopic cross-section database of step S1, according to the historical temperature distribution and historical density distribution of the fuel and reflector layer, a bilinear interpolation method is used to calculate and obtain the current microscopic reaction cross section of each nuclide;

[0085] According to the historical density distribution and the historical node displacement distribution, the current nucleon density is obtained;

[0086] The neutron macroscopic reaction cross section of the fuel and reflector materials is calculated based on the current microscopic reaction cross section and current nucleon density of each nuclide. The solution formula for the neutron macroscopic reaction cross section is:

[0087]

[0088] Where Σ is the neutron macroscopic cross section, N k is the current nuclear density of the kth nuclide; σ k is the current microscopic reaction cross section of the kth nuclide.

[0089] B2. Perform neutron transport kinetics calculation based on the neutron macroscopic reaction cross section obtained in step B1 and the corresponding neutron transport kinetics calculation grid in the history grid to obtain the core power density distribution; wherein the neutron transport kinetics calculation formula is:

[0090]

[0091] in, is the neutron angular flux at position r, with energy E, moving direction Ω, and time t; ∑(r,E) is the neutron macroscopic reaction cross section at position r, with energy E; q s (r,Ω,E,t) is the scattered neutron source at position r, with energy E, moving direction Ω and time t; q f (r,Ω,E,t) is the fission neutron source at position r, energy E, direction Ω, and time t. is the partial differential operator, and ▽ is the divergence operator.

[0092] B3. Using the interpolation method, calculate the heat source term required for the corresponding thermal hydraulic calculation based on the core power density distribution obtained in step B2;

[0093] B4. Based on the core power density distribution obtained in step B2, the heat source term obtained in step B3, and the corresponding thermal-hydraulic calculation grid in the historical grid, thermal-hydraulic calculations are performed on the fuel and reflector layer in the core based on the physical property parameters obtained in the thermal-hydraulic calculations to obtain the current temperature distribution, current density distribution, and current pressure distribution of the core; the current temperature distribution and current density distribution are respectively updated as the historical temperature distribution and historical density distribution of the next iteration; wherein, in step B4, the formula for the thermal-hydraulic calculation is:

[0094]

[0095]

[0096]

[0097]

[0098] Among them, ρ L is the fluid density, u L is the fluid velocity, p is the pressure, τ is the viscous stress tensor, f L is the generalized momentum source term, h is the fluid specific enthalpy, k L is the thermal conductivity of the fluid, T L is the fluid temperature, T s is the fluid temperature, ρ s is the solid density, c p is the specific heat capacity, k s is the solid thermal conductivity, Q is the heat source term;

[0099] B5. Based on the current temperature distribution, current pressure distribution, current external force load, and initial mechanical calculation grid of the core in step B4, and based on the physical parameters obtained from the mechanical calculation, perform mechanical calculations on the core and reflector layer, obtain the current node displacements of the core and reflector layer on the calculation grid, and update them as the historical node displacements for the next iteration. The updated historical grid for the next iteration is obtained using the updated historical node displacements. The mechanical calculation formula corresponding to the mechanical calculations of the core and reflector layer is:

[0100]

[0101]

[0102] σ=D:ε

[0103] u s is the node displacement, σ is the stress tensor, f s is the external load, Tr represents the transpose, D is the elastic matrix, and ε is the strain tensor.

[0104] The second termination condition is: determining whether the power density distribution, temperature distribution, density distribution, and node displacement of the core have converged at the current time; if the set convergence limit is not met, returning to step B1 until the set convergence limit is met; if the set convergence limit is met, executing step A3; wherein the convergence determination formula is specifically:

[0105]

[0106]

[0107]

[0108]

[0109]

[0110]

[0111]

[0112]

[0113] in, are the maximum errors of power density, liquid temperature, solid temperature, liquid density, solid density, x-direction node displacement, y-direction node displacement, and z-direction node displacement; ε Q 、 are the maximum allowable errors of power density, liquid temperature, solid temperature, liquid density, solid density, and nodal displacement, respectively, that is, the convergence limits corresponding to power density, liquid temperature, solid temperature, liquid density, solid density, and nodal displacement, respectively; n+1 is the n+1th time step, i is the i-th iteration, and i+1 is the i+1th iteration.

[0114] This embodiment also provides a computer program product, which may be, but is not limited to, a computer-readable storage medium or a computer device.

[0115] A computer program is stored on a computer-readable storage medium, and when the computer program is executed by a processor, the steps of the above-mentioned fast spectrum small nuclear reactor core transient multi-physics field coupling method are implemented. The computer-readable storage medium can be a readable signal medium or a readable storage medium. The readable storage medium can be, for example, but not limited to, an electrical, magnetic, optical, electromagnetic, infrared, or semiconductor system, device or device, or any combination thereof. More specific examples of readable storage media (a non-exhaustive list) include: an electrical connection with one or more wires, a portable disk, a hard disk, a random access memory (RAM), a read-only memory (ROM), an erasable programmable read-only memory (EPROM or flash memory), an optical fiber, a portable compact disk read-only memory (CD-ROM), an optical storage device, a magnetic storage device, or any suitable combination of the above.

[0116] The above method forms a computer program product, so that the above method can be applied to a computer device, which includes a processor, a memory connected to the processor, and a computer program that can be run on the processor. When the processor executes the computer program, the steps of the above-mentioned fast spectrum small nuclear reactor core transient multi-physical field coupling method are implemented; the computer device here can be a computer, notebook, PDA, various cloud servers and other computing devices, and the processor can be a general-purpose processor, a digital signal processor, an application-specific integrated circuit or other programmable logic device, etc.

Claims

1. A fast spectrum small nuclear reactor core transient multi-physics field coupling method, characterized in that: The following steps are involved: S1. Construct a neutron microscopic cross-section database for nuclear reactor cores; At the same time, a neutron transport mechanics calculation grid model, a thermal hydraulic calculation grid model, and a mechanical calculation grid model are constructed; the neutron transport mechanics calculation grid model is defined as the initial neutron transport mechanics calculation grid, the thermal hydraulic calculation grid model is defined as the initial thermal hydraulic calculation grid, and the mechanical calculation grid model is defined as the initial mechanical calculation grid; At the same time, the physical parameters of thermal hydraulic calculation and mechanical calculation are set based on the material of the reactor core; S2. Setting a preset total simulation time and determining initial conditions and boundary conditions for calculation based on the reactor's initial time operating state; the initial conditions include initial temperature distribution, initial density distribution, and initial node displacement distribution; S3. Time stepping to update the current time; based on the neutron microscopic cross-section database, the neutron transport dynamics calculation grid model, the thermal-hydraulic calculation grid model, the mechanical calculation grid model, the physical property parameters of the thermal-hydraulic calculation, and the physical property parameters of the mechanical calculation in step S1, the power density distribution, temperature distribution, density distribution of the core, and the displacement of each node of the core and reflector calculation grid at the current time are calculated, and iteration is performed until the power density distribution, temperature distribution, density distribution of the core, and the displacement of each node of the core and reflector calculation grid meet the set convergence limit, then it is determined whether the final time of the preset total simulation time has been reached; if the set final time has been reached, the calculation ends; If the final time has not been reached, the time stepping continues and the current time is updated until the set final time is reached. The calculation ends and the transient characteristics of the nuclear reactor core within the preset total simulation time are obtained; Step S3 is specifically as follows: S3.

1. Obtain historical temperature distribution, where the historical temperature is the initial temperature distribution; obtain historical density distribution, where the historical density distribution is the initial density distribution; obtain historical grids, where the historical grids are the initial neutron transport kinetics calculation grid and the initial thermal hydraulic calculation grid; obtain historical node displacement distribution, where the historical node displacement distribution is the initial node displacement distribution; S3.

2. Execute the first loop system until the first end condition is met; The first circulation system comprises: A1. Time stepping to obtain the current time; A2. Execute the second loop system until the second end condition is met; A3. Determine whether to execute the next time step; The first end condition is: determining whether the current time is the end time of the preset total simulation time, if so, the calculation ends; if not, returning to step A1 until the calculation ends and the transient characteristics of the nuclear reactor core within the preset total simulation time are obtained; The second circulation system comprises: B1. Based on the neutron microscopic cross-section database of step S1, according to the historical temperature distribution and historical density distribution of the fuel and reflector layer, a bilinear interpolation method is used to calculate and obtain the current microscopic reaction cross section of each nuclide; According to the historical density distribution and the historical node displacement distribution, the current nucleon density is obtained; Calculate the neutron macroscopic reaction cross section of the fuel and reflector corresponding materials based on the current microscopic reaction cross section and current nucleon density of each nuclide; B2. Perform neutron transport kinetics calculations based on the neutron macroscopic reaction cross sections of the fuel and reflector materials obtained in step B1 and the corresponding neutron transport kinetics calculation grid in the history grid to obtain the core power density distribution; B3. Using the interpolation method, calculate the heat source term Q required for the corresponding thermal hydraulic calculation based on the core power density distribution obtained in step B2; B4. Based on the core power density distribution obtained in step B2, the heat source term Q obtained in step B3, and the corresponding thermal-hydraulic calculation grid in the historical grid, and based on the physical property parameters calculated by the thermal-hydraulic calculation, thermal-hydraulic calculations are performed on the fuel and reflector layer in the core to obtain the current temperature distribution, current density distribution, and current pressure distribution of the core; the current temperature distribution and current density distribution are respectively updated as the historical temperature distribution and historical density distribution of the next iteration; B5. Based on the current temperature distribution, current pressure distribution, current external force load, and initial mechanical calculation grid of the core in step B4, and based on the physical property parameters obtained from the mechanical calculation, mechanical calculations are performed on the core and reflector to obtain the current node displacements of the core and reflector on the calculation grid. These displacements are updated as the historical node displacements for the next iteration, and the historical grid for the next iteration is obtained using the updated historical node displacements. The second termination condition is: judging whether the power density distribution, temperature distribution, density distribution and node displacement of the core have converged at the current time; if the set convergence limit is not met, returning to step B1 until the set convergence limit is met; if the set convergence limit is met, executing step A3.

2. The fast spectrum small nuclear reactor core transient multi-physics field coupling method according to claim 1, characterized in that: Step S1 is specifically as follows: A Monte Carlo program is used to construct a neutron micro-cross-section database consisting of corresponding neutron micro-cross-sections at different state points based on the state parameters of fuel temperature, fuel density, coolant temperature, coolant density, reflector temperature and reflector density at different state points through the evaluation database; According to the geometry and material of the reactor core, the neutron transport kinetics calculation grid model, the thermal hydraulic calculation grid model and the mechanical calculation grid model are constructed based on the meshing software.

3. The fast spectrum small nuclear reactor core transient multi-physics field coupling method according to claim 2, characterized in that: In step B1, the solution formula for the neutron macroscopic reaction cross section is: Where Σ is the neutron macroscopic cross section, N k is the current nuclear density of the kth nuclide; σ k is the current microscopic reaction cross section of the kth nuclide.

4. The fast spectrum small nuclear reactor core transient multi-physics field coupling method according to claim 3, characterized in that: In step B2, the neutron transport kinetics calculation formula is: in, is the neutron angular flux at position r, with energy E, moving direction Ω, and time t; Σ(r,E) is the neutron macroscopic reaction cross section at position r, with energy E; q s (r,Ω,E,t) is the scattered neutron source at position r, with energy E, moving direction Ω and time t; q f (r,Ω,E,t) is the fission neutron source at position r, energy E, direction Ω, and time t.

5. The fast spectrum small nuclear reactor core transient multi-physics field coupling method according to claim 4, characterized in that: In step B4, the thermal hydraulic calculation formula is: Among them, ρ L is the fluid density, u L is the fluid velocity, p is the pressure, τ is the viscous stress tensor, f L is the generalized momentum source term, h is the fluid specific enthalpy, k L is the thermal conductivity of the fluid, T L is the fluid temperature, T s is the fluid temperature, ρ s is the solid density, c p is the specific heat capacity, ρ s is the solid density, k s is the solid thermal conductivity, Q is the heat source term; In step B5, the mechanical calculation formula for the core and reflector is: σ=D:ε u s is the node displacement, σ is the stress tensor, f s is the external load, Tr represents the transpose, D is the elastic matrix; ε is the strain tensor.

6. The fast spectrum small nuclear reactor core transient multi-physics field coupling method according to claim 5, characterized in that: In step B5, the convergence judgment formula is specifically: in, are the maximum errors of power density, liquid temperature, solid temperature, liquid density, solid density, x-direction node displacement, y-direction node displacement, and z-direction node displacement; ε Q 、 ε Ts 、 ε ρs , ε us are the maximum allowable errors of power density, liquid temperature, solid temperature, liquid density, solid density, and node displacement respectively; n+1 is the n+1th time step, i is the i-th iteration, and i+1 is the i+1th iteration.

7. The fast spectrum small nuclear reactor core transient multi-physics field coupling method according to claim 6, characterized in that: In step S1, the Monte Carlo program is OpenMC or MCNP; The evaluation database is the China Evaluation Nuclear Database CENDL or the United States Evaluation Nuclear Database ENDF; The mesh generation software is Gmesh, ICEM or Pointwise.

8. A computer program product having a computer program stored thereon, characterized in that: When the computer program is executed by a processor, the steps of the fast spectrum small nuclear reactor core transient multi-physical field coupling method according to any one of claims 1 to 7 are implemented.

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