A finite element-based multi-physics coupling dynamic response analysis method for nuclear reactor systems

Abstract

A kind of nuclear reactor system multi-physical coupling dynamic response analysis method based on finite element, steps as follows:1, establish nuclear reactor thermal, structural mechanics and water chemistry key model, including reactor system two-fluid six equation model, reactor core sub-channel two-fluid eight equation model, fuel element pellet nonlinear constitutive model and activated corrosion migration model;2, establish reactor neutron physical transport model;3, by traditional finite element and discontinuous finite element method, nuclear reactor system thermal, reactor core sub-channel, structural mechanics, water chemistry analysis and neutron physical model are dispersed, form thermal-power-chemical full coupling, and tightly coupled with neutron physics Large residual equation;4, using transient nonlinear process solver, large residual equation is solved, and the transient distribution of each physical field parameter in the reactor is obtained.The method can analyze the multi-level coupling effect mechanism between complex physical fields in the reactor, and provide technical guidance for the challenging problems that the reactor core may face.

Description

Technical Field

[0001] This invention belongs to the field of multi-physics coupling technology of nuclear reactor systems, specifically involving a finite element method for analyzing the dynamic response of multi-physics coupling in nuclear reactor systems, so as to analyze the coupling mechanism of key physical fields under reactor operation and accident transients. Background Technology

[0002] Nuclear reactor systems are complex and operate in harsh environments, exhibiting intricate multiphysics coupling phenomena. This leads to challenging problems for safe reactor operation, including fouling, component deformation and blockage, fuel rod-grid erosion, pellet-cladding interactions, and reactivity introduction. Currently, traditional single-physics software struggles to analyze these problems, and efficient multiphysics coupling solution methods are insufficient, with the mechanisms of multi-scale multiphysics coupling remaining unclear.

[0003] To address the challenging problem of multiphysics coupling during nuclear reactor operation, a series of numerical reactor research and development projects have been launched internationally, focusing on the complex multiphysics and multiscale coupling problems within the reactor core. Early international research strategies for reactor numerical simulation involved integrating existing software and performing coupling analysis between some computational modules. Currently, to achieve high-fidelity multiphysics coupling analysis of reactor systems, research institutions are gradually developing multiphysics coupling systems that tightly couple all physical modules, including physics, thermodynamics, structure, and chemistry. However, research on multiphysics coupling analysis technology in my country is still mainly limited to software integration and coupling of some physical modules.

[0004] Therefore, it is urgent to establish a multi-physics coupled dynamic response analysis technology for nuclear reactor systems to enhance my country's high-fidelity numerical analysis capabilities for nuclear reactors, solve key challenging problems, and improve the safety and economy of nuclear reactor operation. Summary of the Invention

[0005] To address the aforementioned problems, this invention provides a finite element method for analyzing the dynamic response of multi-physics coupling in nuclear reactor systems. This method first establishes key models of nuclear reactor thermodynamics, structural mechanics, water chemistry, and neutron physics. Then, it discretizes the nuclear-thermal-mechanical-chemical coupling model using the finite element method to obtain a large set of residual equations. Finally, it uses a transient nonlinear process solver to solve the large set of residual equations, obtaining the transient distribution of various physical field parameters within the reactor. This allows for the analysis of the multi-level coupling effect mechanism between complex physical fields within the reactor, solving potentially challenging problems faced by the reactor core, and revealing the coupling mechanism of key physical fields during reactor operation and accident transients.

[0006] To achieve the above objectives, the present invention adopts the following technical solution:

[0007] A multi-physics coupled dynamic response analysis method for nuclear reactor systems based on the finite element method is proposed, with the following steps:

[0008] Step 1: For the complex thermal, mechanical, and water chemical phenomena involved in the nuclear reactor system, establish a fully coupled model of the nuclear reactor system's thermal-core sub-channels-structural mechanics-water chemistry. Outside the core, use a coarse mesh to establish a two-fluid, six-equation thermal-hydraulic model and key equipment component models of the nuclear reactor system, possessing complete modeling and calculation capabilities for the reactor's primary loop system, forming a reactor system thermal analysis module. The two-fluid, six-equation model consists of the liquid / vapor phase mass conservation equation, the liquid / vapor phase momentum conservation equation, and the liquid / vapor phase energy conservation equation. The two-fluid, six-equation thermal-hydraulic model is as follows:

[0009]

[0010] In the formula:

[0011] α l —Liquid phase cavitation fraction;

[0012] ρ l —Liquid phase density;

[0013] u l —Liquid phase velocity;

[0014] t — time;

[0015] z — Spatial position in the z-direction;

[0016] Γ g —The energy transferred between phases;

[0017] α g —Vacuum phase cavitation fraction;

[0018] ρ g —Vapor phase density;

[0019] u g —Vapor phase velocity;

[0020] p – pressure;

[0021] g z — Gravitational acceleration; F int —Interphase frictional resistance;

[0022] F wall,l —Frictional resistance of the liquid phase wall; u int —Interface speed;

[0023] F wall,g —Frictional resistance of the vapor phase wall;

[0024] e l —Liquid phase internal energy;

[0025] Q wl —Heat transferred from the wall to the liquid phase;

[0026] Q il —Heat transferred between the interface and the liquid phase;

[0027] Γ ig —Mass transfer between phase bubbles;

[0028] —The liquid phase enthalpy of mass transfer between interfaces;

[0029] Γ w —The rate of vapor generation during wall boiling / condensation;

[0030] h l —The liquid phase enthalpy caused by mass transfer due to vapor generation at the wall surface;

[0031] e g —Vacuum phase internal energy;

[0032] Q wg —Heat transferred from the wall to the vapor phase;

[0033] Q ig —The heat transferred between the interface and the vapor phase;

[0034] —Vapor phase enthalpy of mass transfer between interfaces;

[0035] h' g —The vapor phase enthalpy caused by mass transfer due to vapor generation on the wall surface;

[0036] A two-fluid eight-equation thermo-hydraulic model, a turbulence mixing model, a phase interface heat transfer model, a lattice drag model, a phase interface drag model, and a critical heat flux density model are established for the core sub-channels, forming a thermal analysis module for the core sub-channels. The two-fluid eight equations consist of the liquid / vapor phase mass conservation equation, the liquid / vapor phase momentum conservation equation, the liquid / vapor phase energy conservation equation, and the liquid / vapor phase axial momentum conservation equation. The two-fluid eight-equation thermo-hydraulic model is as follows:

[0037]

[0038] In the formula:

[0039] Subscript l — liquid phase;

[0040] Subscript g – vapor phase;

[0041] A – Subchannel area;

[0042] C – Mass exchange caused by transverse turbulent mixing and cavitation drift between channels; S – Gap width;

[0043] U—Axial velocity;

[0044] V—Transverse velocity of fluid in the channel gap;

[0045] z—Axial length of the channel;

[0046] Γ—phase change flow rate per unit volume;

[0047] α — Volume fraction;

[0048] ε—Crossflow direction indicator;

[0049] ρ — density;

[0050] D h —Equivalent diameter of the channel;

[0051] K—Local resistance coefficient;

[0052] K I —Interfacial drag coefficient;

[0053] M—Momentum exchange caused by transverse turbulent mixing and cavitation drift between channels; —Axial velocity of the fluid participating in the phase transition;

[0054] f w —Road friction coefficient;

[0055] g — acceleration due to gravity;

[0056] p – pressure;

[0057] K G — Gap resistance friction coefficient;

[0058] —The transverse velocity of the phase change fluid;

[0059] l gap — The mixing length between adjacent transverse channels, generally taken as the center-to-center distance between the two channels; E — Energy exchange caused by transverse turbulent mixing and cavitation drift between channels;

[0060] P w —Channel wet perimeter;

[0061] h — enthalpy;

[0062] q I —The amount of heat transferred between a unit volume of fluid and the phase interface;

[0063] q — heat flux density of the heated wall surface;

[0064] Φ mn—The heating perimeter share corresponding to channel m;

[0065] η wv —The heat share of the vapor phase directly heated by the wall;

[0066] A nonlinear thermal conductivity and mechanical constitutive model of the fuel element under irradiation thermo-coupling conditions was established, including models for elastoplasticity, high-temperature creep, high-temperature phase transition, high-temperature oxidation, and cladding failure, forming a mechanical performance analysis module. The thermal conductivity and mechanical constitutive model is as follows:

[0067]

[0068] In the formula:

[0069] ρ — fuel density;

[0070] c p —Specific heat capacity of fuel at constant pressure;

[0071] T—Fuel element temperature;

[0072] t — time;

[0073] k—thermal conductivity;

[0074] Q f —Fuel heat release rate;

[0075] σ—Cauchy stress tensor;

[0076] f—volume force;

[0077] A corrosion product deposition model based on a comprehensive conduction model was established, focusing on key hydrochemical phenomena related to the diffusion of critical nuclides, corrosion, and deposition of corrosion products within the reactor. This resulted in a hydrochemical analysis module. The corrosion product deposition model is as follows:

[0078]

[0079] In the formula:

[0080] I—Metal quality of corrosion products on the reactor core surface;

[0081] t — time;

[0082] V SD —Deposition rate of dissolved corrosion products;

[0083] V PD —Deposition rate of particulate corrosion products;

[0084] V EC —Erosion rate of sedimentary layers;

[0085] Step 2: Establish a neutron physics transport model. Unstructured grid discretization is used for refined modeling. The accuracy of angle and energy processing is improved by combining the discrete ordinate method and the multi-group energy approximation method. Simultaneously, based on resonant self-shield calculation technology, the effects of resonant interference, temperature distribution, spatial self-shield, resonant elastic scattering, edge effects, and multi-group equivalent effects are comprehensively considered. Combined with the multi-group database, continuous energy database, and burnup database of the target reactor type, a core physics module based on deterministic methods is formed. The neutron physics transport model is as follows:

[0086]

[0087] In the formula:

[0088] v — neutron velocity;

[0089] t — time;

[0090] Ω – Direction;

[0091] E – Energy;

[0092] f—scattering phase function;

[0093] ∑ t —Total reaction cross section;

[0094] ∑ s —Scattering cross section;

[0095] ∑ f — Fission cross section;

[0096] r — position;

[0097] φ — Neutron angular flux density;

[0098] ν—average fission neutron number;

[0099] χ—fission spectral function;

[0100] S e —Additional neutron source;

[0101] Step 3: Establish a data mapping between the neutron physics field and the thermo-mechanical-chemical field based on the nearest point transfer and L2 mapping algorithm; then, discretize the fully coupled model of the nuclear reactor system's thermo-engineering, core sub-channel, structural mechanics, and water chemistry using the finite element method (discrete the solid domain using the traditional finite element method and the fluid domain using the discontinuous finite element method) to obtain the weak solution form of the model, and construct a large-scale residual equation system that is fully coupled with thermo-mechanical-chemical fields and tightly coupled with neutron physics.

[0102] Step 4: Under a unified framework, the large residual equation system is iteratively solved using the preprocessing JFNK method with transient nonlinear solution tools to obtain the temporal and spatial distribution of physical quantities in the nuclear-thermal-mechanical-chemical field.

[0103] Compared with the prior art, the present invention has the following advantages and beneficial effects:

[0104] 1. This method uses full coupling technology to achieve deep coupling of thermo-mechanical-chemical multi-scale multi-physics fields, which solves the prominent problems of low numerical discretization accuracy, incomplete consideration of coupling effects, and low solution efficiency of traditional analysis methods;

[0105] 2. The tight coupling technique between the neutron physics field and the fully coupled thermo-mechanical-chemical field in this method breaks through the technical bottleneck of insufficient coupling depth in the traditional single tight coupling method, and also solves the problems of poor convergence and huge computational consumption of the fully coupled method.

[0106] 3. This method takes into account the impact of multiphysics coupling effects on nuclear reactor operation and accident processes, and can provide a reference for the optimized design, safe operation and severe accident safety strategies of nuclear reactors. Attached Figure Description

[0107] Figure 1 This is a flowchart of the multi-physics coupling dynamic response analysis method for nuclear reactor systems based on the finite element method of the present invention.

[0108] Figure 2 This is a schematic diagram of a fully coupled model of the system's thermal-core sub-channels-structural mechanics-water chemistry.

[0109] Figure 3 This is a schematic diagram of constructing a large set of residual equations based on a unified finite element framework. Detailed Implementation

[0110] The present invention will now be described in further detail with reference to the accompanying drawings and specific embodiments.

[0111] like Figure 1 As shown, this invention provides a multi-physics coupled dynamic response analysis method for nuclear reactor systems based on the finite element method. The specific method is as follows:

[0112] Step 1: If the finite element platform MOOSE is used, a fully coupled model of the nuclear reactor system's thermal-core sub-channels-structural mechanics-water chemistry needs to be established in the Kernel section, such as... Figure 2As shown; a coarse-grid model of the nuclear reactor system and key equipment components are established outside the reactor core using a two-fluid, six-equation thermal-hydraulic model. This model possesses complete modeling and calculation capabilities for the reactor's primary loop system, forming a reactor system thermal analysis module. The two-fluid, six-equation model consists of the liquid / vapor phase mass conservation equation, the liquid / vapor phase momentum conservation equation, and the liquid / vapor phase energy conservation equation. The two-fluid, six-equation thermal-hydraulic model is as follows:

[0113]

[0114] In the formula:

[0115] α l —Liquid phase cavitation fraction;

[0116] ρ l —Liquid phase density;

[0117] u l —Liquid phase velocity;

[0118] t — time;

[0119] z — Spatial position in the z-direction;

[0120] Γ g —The energy transferred between phases;

[0121] α g —Vacuum phase cavitation fraction;

[0122] ρ g —Vapor phase density;

[0123] u g —Vapor phase velocity;

[0124] p – pressure;

[0125] g z —Acceleration due to gravity;

[0126] F int —Interphase frictional resistance;

[0127] F wall,l —Frictional resistance of the liquid phase wall;

[0128] u int —Interface speed;

[0129] F wall g—frictional resistance of the vapor phase wall;

[0130] e l —Liquid phase internal energy;

[0131] Q wl —The heat transferred from the wall to the liquid phase; Q il—Heat transferred between the interface and the liquid phase;

[0132] Γ ig —Mass transfer between phase bubbles;

[0133] —The liquid phase enthalpy of mass transfer between interfaces;

[0134] Γ w —The rate of vapor generation during wall boiling / condensation;

[0135] h l —The liquid phase enthalpy caused by mass transfer due to vapor generation at the wall surface;

[0136] e g —Vacuum phase internal energy;

[0137] Q wg —Heat transferred from the wall to the vapor phase;

[0138] Q ig —The heat transferred between the interface and the vapor phase;

[0139] —Vapor phase enthalpy of mass transfer between interfaces;

[0140] h' g —The vapor phase enthalpy caused by mass transfer due to vapor generation on the wall surface;

[0141] A two-fluid eight-equation thermo-hydraulic model, a turbulence mixing model, a phase interface heat transfer model, a lattice drag model, a phase interface drag model, and a critical heat flux density model are established for the core sub-channels, forming a thermal analysis module for the core sub-channels. The two-fluid eight equations consist of the liquid / vapor phase mass conservation equation, the liquid / vapor phase momentum conservation equation, the liquid / vapor phase energy conservation equation, and the liquid / vapor phase axial momentum conservation equation. The two-fluid eight-equation thermo-hydraulic model is as follows:

[0142]

[0143] In the formula:

[0144] Subscript l — liquid phase;

[0145] Subscript g – vapor phase;

[0146] A – Subchannel area;

[0147] C – Mass exchange caused by transverse turbulent mixing and cavitation drift between channels; S – Gap width;

[0148] U—Axial velocity;

[0149] V—Transverse velocity of fluid in the channel gap;

[0150] z—Axial length of the channel;

[0151] Γ—phase change flow rate per unit volume;

[0152] α — Volume fraction;

[0153] ε—Crossflow direction indicator;

[0154] ρ — density;

[0155] D h —Equivalent diameter of the channel;

[0156] K—Local resistance coefficient;

[0157] K I —Interfacial drag coefficient;

[0158] M—Momentum exchange caused by transverse turbulent mixing and cavitation drift between channels; —Axial velocity of the fluid participating in the phase transition;

[0159] f w —Road friction coefficient;

[0160] g — acceleration due to gravity;

[0161] p – pressure;

[0162] K G — Gap resistance friction coefficient;

[0163] —The transverse velocity of the phase change fluid;

[0164] l gap — The mixing length between adjacent transverse channels, generally taken as the center-to-center distance between the two channels; E — Energy exchange caused by transverse turbulent mixing and cavitation drift between channels;

[0165] P w —Channel wet perimeter;

[0166] h — enthalpy;

[0167] q I —The amount of heat transferred between a unit volume of fluid and the phase interface;

[0168] q — heat flux density of the heated wall surface;

[0169] Φ mn —The heating perimeter share corresponding to channel m;

[0170] η wv —The heat share of the vapor phase directly heated by the wall;

[0171] A nonlinear thermal conductivity and mechanical constitutive model of the fuel element under irradiation thermo-coupling conditions was established, including models for elastoplasticity, high-temperature creep, high-temperature phase transition, high-temperature oxidation, and cladding failure, forming a mechanical performance analysis module. The thermal conductivity and mechanical constitutive model is as follows:

[0172]

[0173] In the formula:

[0174] ρ — fuel density;

[0175] c p —Specific heat capacity of fuel at constant pressure;

[0176] T—Fuel element temperature;

[0177] t — time;

[0178] k—thermal conductivity;

[0179] Q f —Fuel heat release rate;

[0180] σ—Cauchy stress tensor;

[0181] f—volume force;

[0182] A corrosion product deposition model based on a comprehensive conduction model was established, focusing on key hydrochemical phenomena related to the diffusion of critical nuclides, corrosion, and deposition of corrosion products within the reactor. This resulted in a hydrochemical analysis module. The corrosion product deposition model is as follows:

[0183]

[0184] In the formula:

[0185] I—Metal quality of corrosion products on the reactor core surface;

[0186] t — time;

[0187] V SD —Deposition rate of dissolved corrosion products;

[0188] V PD —Deposition rate of particulate corrosion products;

[0189] V EC —Erosion rate of sedimentary layers;

[0190] Step 2: If the finite element platform MOOSE is used, a neutron physics transport model needs to be established in the kernel section. Unstructured mesh discretization is achieved by using the preprocessing software Trelis to achieve fine modeling. The accuracy of angle and energy processing is improved by combining the discrete ordinate method and the multi-group energy approximation method. At the same time, the influence of resonance interference effect, temperature distribution effect, spatial self-screen effect, resonance elastic scattering effect, edge effect and multi-group equivalent effect are comprehensively considered based on the resonance self-screen calculation technology. Combined with the multi-group database, continuous energy database and burnup database of the target reactor type, a core physics module based on deterministic method is formed. The neutron physics transport model is as follows:

[0191]

[0192] In the formula:

[0193] v — neutron velocity;

[0194] t — time;

[0195] Ω – Direction;

[0196] E – Energy;

[0197] f—scattering phase function;

[0198] ∑ t —Total reaction cross section;

[0199] ∑ s —Scattering cross section;

[0200] ∑ f — Fission cross section;

[0201] r — position;

[0202] φ — Neutron angular flux density;

[0203] ν—average fission neutron number;

[0204] χ—fission spectral function;

[0205] S e —Additional neutron source;

[0206] Step 3: If the finite element platform MOOSE is used, a data mapping between the neutron physics field and the thermo-mechanical-chemical field can be established in Util; then, as follows... Figure 3 As shown, the fully coupled thermal-core subchannel-structural mechanics-water chemistry model of the nuclear reactor system is discretized using the finite element method (the solid domain is discretized using the traditional finite element method, and the fluid domain is discretized using the discontinuous finite element method) to obtain the weak solution form of the model, and a large set of residual equations that are fully coupled with thermo-mechanical-chemical physics and tightly coupled with neutron physics is constructed.

[0207] Step 4: If the finite element platform MOOSE is used for calculation, the preprocessing JFNK method can be used to iteratively solve the large residual equation system, and the distribution of physical quantities in the nuclear-thermal-mechanical-chemical field in time and space can be obtained through Postprocessors.

[0208] The above description is a further detailed explanation of the present invention in conjunction with specific preferred embodiments. It should not be considered that the specific embodiments of the present invention are limited to this. For those skilled in the art, several simple deductions or substitutions can be made without departing from the concept of the present invention, and all such deductions or substitutions should be considered to fall within the scope of patent protection determined by the submitted claims.

Claims

1. A method for analyzing the multi-physics coupled dynamic response of a nuclear reactor system based on the finite element method, characterized in that: The steps are as follows: Step 1: For the complex thermal, mechanical, and water chemical phenomena involved in the nuclear reactor system, establish a fully coupled model of the nuclear reactor system's thermal-core sub-channels-structural mechanics-water chemistry. Outside the core, use a coarse mesh to establish a two-fluid, six-equation thermal-hydraulic model and key equipment component models of the nuclear reactor system, possessing complete modeling and calculation capabilities for the reactor's primary loop system, forming a reactor system thermal analysis module. The two-fluid, six-equation model consists of the liquid / vapor phase mass conservation equation, the liquid / vapor phase momentum conservation equation, and the liquid / vapor phase energy conservation equation. The two-fluid, six-equation thermal-hydraulic model is as follows: In the formula: —Liquid phase cavitation fraction; —Liquid phase density; —Liquid phase velocity; --time; —Position in the z-direction of space; —The energy transferred between phases; —Vacuum phase cavitation fraction; —Vapor phase density; —Vapor phase velocity; --pressure; —Acceleration due to gravity; —Interphase frictional resistance; —Frictional resistance of the liquid phase wall; —Interface speed; —Frictional resistance of the vapor phase wall; —Liquid phase internal energy; —Heat transferred from the wall to the liquid phase; —Heat transferred between the interface and the liquid phase; —Mass transfer between phase bubbles; —The liquid phase enthalpy of mass transfer between interfaces; —The rate of vapor generation during wall boiling / condensation; —The liquid phase enthalpy caused by mass transfer due to vapor generation on the wall surface; —Vacuum phase internal energy; —Heat transferred from the wall to the vapor phase; —The heat transferred between the interface and the vapor phase; —Vapor phase enthalpy of mass transfer between interfaces; —The vapor phase enthalpy caused by mass transfer due to vapor generation on the wall surface; A two-fluid eight-equation thermo-hydraulic model, a turbulence mixing model, a phase interface heat transfer model, a lattice drag model, a phase interface drag model, and a critical heat flux density model are established for the core sub-channels, forming a thermal analysis module for the core sub-channels. The two-fluid eight equations consist of the liquid / vapor phase mass conservation equation, the liquid / vapor phase momentum conservation equation, the liquid / vapor phase energy conservation equation, and the liquid / vapor phase axial momentum conservation equation. The two-fluid eight-equation thermo-hydraulic model is as follows: In the formula: Subscript —Liquid phase; Subscript —Vacuum phase; —Subchannel area; —Mass exchange caused by lateral turbulent mixing and cavitation drift between channels; —Gap width; —Axial velocity; — Lateral velocity of fluid in the channel gap; —Axial length of the channel; —Phase change flow rate per unit volume; —Volume share; —Crossflow direction indicator; --density; —Equivalent diameter of the channel; —Local drag coefficient; —Interfacial drag coefficient; —Momentum exchange caused by lateral turbulent mixing and cavitation drift between channels; —Axial velocity of the fluid participating in the phase transition; —Road friction coefficient; —Acceleration due to gravity; --pressure; — Gap resistance friction coefficient; —The transverse velocity of the phase change fluid; —The overlap length between adjacent transverse channels is generally taken as the center-to-center distance between the two channels; —Energy exchange caused by lateral turbulent mixing and cavitation drift between channels; —Channel wet perimeter; Enthalpy; —The amount of heat transferred between a unit volume of fluid and the phase interface; —Heat flux density of the heated wall surface; —The heating perimeter share corresponding to channel m; —The heat share of the vapor phase directly heated by the wall; A nonlinear thermal conductivity and mechanical constitutive model of the fuel element under irradiation thermo-coupling conditions was established, including models for elastoplasticity, high-temperature creep, high-temperature phase transition, high-temperature oxidation, and cladding failure, forming a mechanical performance analysis module. The thermal conductivity and mechanical constitutive model is as follows: In the formula: —Fuel density; —Specific heat capacity of fuel at constant pressure; —Fuel element temperature; --time; — Thermal conductivity; —Fuel heat release rate; —Cauchy stress tensor; —Volume force; A corrosion product deposition model based on a comprehensive conduction model was established, focusing on key hydrochemical phenomena related to the diffusion of critical nuclides, corrosion, and deposition of corrosion products within the reactor. This resulted in a hydrochemical analysis module. The corrosion product deposition model is as follows: In the formula: —Quality of corrosion products on the reactor core surface; --time; —Deposition rate of dissolved corrosion products; —Deposition rate of particulate corrosion products; —Erosion rate of sedimentary layers; Step 2: Establish a neutron physics transport model. Unstructured grid discretization is used for refined modeling. The accuracy of angle and energy processing is improved by combining the discrete ordinate method and the multi-group energy approximation method. Simultaneously, based on resonant self-shield calculation technology, the effects of resonant interference, temperature distribution, spatial self-shield, resonant elastic scattering, edge effects, and multi-group equivalent effects are comprehensively considered. Combined with the multi-group database, continuous energy database, and burnup database of the target reactor type, a core physics module based on deterministic methods is formed. The neutron physics transport model is as follows: In the formula: — Neutron speed; --time; --direction; --energy; —Scattering phase function; —Total reaction cross section; —Scattering cross section; — Fission cross section; --Location; —— Neutron angular flux density; —Average number of fission neutrons; — Fission spectral function; —Additional neutron source; Step 3: Establish a data mapping between the neutron physics field and the thermo-mechanical-chemical field based on the nearest point transfer and L2 mapping algorithm; then, discretize the fully coupled model of the nuclear reactor system's thermo-core sub-channel-structural mechanics-water chemistry using the finite element method to obtain the weak solution form of the model, and construct a large set of residual equations that are fully coupled with thermo-mechanical-chemical fields and tightly coupled with neutron physics. Step 4: Under a unified framework, the large residual equation system is iteratively solved using the preprocessing JFNK method with transient nonlinear solution tools to obtain the temporal and spatial distribution of physical quantities in the nuclear-thermal-mechanical-chemical field.

2. The method for multi-physics coupled dynamic response analysis of a nuclear reactor system based on the finite element method according to claim 1, characterized in that: Step 3 describes discretizing the fully coupled model of the nuclear reactor system's thermal-core subchannel-structural mechanics-water chemistry using the finite element method. Specifically, the solid domain is discretized using the finite element method, and the fluid domain is discretized using the discontinuous finite element method.

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