Lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method and system

Abstract

The invention relates to a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method and system, and the method comprises the steps: constructing a reactor core geometric model, and generating reaction section data; the reactor core geometric model comprises a fuel rod area and a coolant area, and the reaction cross section data comprises multiple groups, multiple temperature points and multiple void fractions; calculating reactor core power distribution based on a reactor core neutronics calculation model according to the reaction cross section data; according to the reactor core power distribution, based on a reactor core gas-liquid two-phase thermal hydraulic calculation model, thermal parameters of the current time step are obtained, and the thermal parameters comprise the fuel rod temperature, the coolant temperature and the void fraction. According to the calculation method provided by the invention, the influence of the gas phase on neutronics characteristics can be accurately simulated, and the reactor core power distribution and thermal parameters in a gas-liquid two-phase flow state can be accurately predicted, so that the power transient safety problem caused by the gas-liquid two-phase flow possibly occurring in the reactor core under a serious accident can be solved; and technical support is provided for safety analysis and accident evaluation of the lead-based reactor.

Description

TECHNICAL FIELD

[0001] The present application relates to the technical field of nuclear reactor design, in particular to a lead-based reactor core gas-liquid two-phase nuclear heat coupling calculation method and system. BACKGROUND

[0002] With the development of nuclear energy technology, as an important type of fourth-generation advanced nuclear energy system, lead-based reactors have attracted widespread attention due to their inherent safety, economy, and sustainability. Nuclear heat coupling calculation of the lead-based reactor core is a key technology for reactor design, safety analysis, and operation optimization, and is of great significance for accurately predicting the neutron and thermal-hydraulic characteristics in the core.

[0003] In the prior art, the nuclear heat coupling calculation in the field of lead-based reactors is limited to the nuclear heat coupling calculation of single-phase flow state of liquid metal. In the severe accident condition of the lead-based reactor, the core may appear gas-liquid two-phase flow state, and the existence of gas phase will significantly affect the neutron characteristics, and then affect the power distribution and thermal-hydraulic characteristics; specifically, two-phase flow will induce core power transient, threatening the safety boundary of the reactor.

[0004] Therefore, it is urgent to develop a gas-liquid two-phase nuclear heat coupling calculation method suitable for lead-based reactors, which can accurately simulate the influence of gas phase on neutron characteristics and accurately predict the core power distribution and thermal parameters under gas-liquid two-phase flow state, and provide technical support for safety analysis and accident evaluation of lead-based reactors. SUMMARY

[0005] The present application provides a lead-based reactor core gas-liquid two-phase nuclear heat coupling calculation method to solve the problem that the existing nuclear heat coupling calculation method of lead-based reactor is only applicable to single-phase flow of liquid metal and cannot cope with the power transient safety problem caused by gas-liquid two-phase flow in the core under severe accident.

[0006] The present application provides a lead-based reactor core gas-liquid two-phase nuclear heat coupling calculation method to solve the technical problem.

[0007] The present application provides a lead-based reactor core gas-liquid two-phase nuclear heat coupling calculation method, which comprises: S100, constructing a core geometry model and generating reaction cross-section data; wherein the core geometry model comprises a fuel rod region and a coolant region, and the reaction cross-section data comprises multiple groups, multiple temperature points and multiple void fractions; S120, calculating the core power distribution based on the core neutron calculation model according to the reaction cross-section data; S140, based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model, the thermal parameters for the current time step are obtained. These thermal parameters include fuel rod temperature, coolant temperature, and void fraction. The core gas-liquid two-phase thermal-hydraulic calculation model includes a thermal-hydraulic model and a transport model. The thermal-hydraulic model is used to calculate the fuel rod temperature and the coolant temperature. The transport model includes gas-liquid two-phase mass conservation equations, gas-liquid two-phase momentum conservation equations, and gas-liquid two-phase energy conservation equations, used to calculate the void fraction. The mass conservation equation for a gas-liquid two-phase system is: ; In the formula, Let be the density of the gas-liquid two-phase mixture. For time, The velocity of the gas and liquid phases, For the fraction of voids, For gas phase density, The density of the liquid phase; The momentum conservation equation for a gas-liquid two-phase system is: ; In the formula, Let be the density of the gas-liquid two-phase mixture. For time, The velocity of the gas and liquid phases, For flow field pressure, The dynamic viscosity of the gas-liquid two-phase mixture is denoted as . For the fraction of voids, The dynamic viscosity of the gas phase. The dynamic viscosity of the liquid phase. It is the acceleration due to gravity; The energy conservation equation for a gas-liquid two-phase system is: ; In the formula, Let be the density of the gas-liquid two-phase mixture. Enthalpy of a gas-liquid two-phase mixture For time, The velocity of the gas and liquid phases, Enthalpy of a gas-liquid two-phase mixture For the fraction of voids, For vapor phase enthalpy, For liquid phase enthalpy, Thermal conductivity, The characteristic temperature of the coolant. This is the rate at which heat is released from fuel to coolant per unit volume.

[0008] Optionally, step S140, which involves obtaining the thermal parameters for the current time step based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model, includes: S142, Based on the core power distribution and the thermal-hydraulic model, the fuel rod temperature and the coolant temperature are calculated. S144, solve the gas-liquid two-phase mass conservation equation, gas-liquid two-phase momentum conservation equation, and gas-liquid two-phase energy conservation equation in sequence to calculate the cavitation fraction.

[0009] Optionally, step S100, the step of constructing the core geometric model and generating reaction section data, includes: S102, Construct the core geometry model; S104, insert bubbles into the coolant region to obtain a core geometry model containing bubbles; this setting is to simulate the gas-liquid two-phase flow state of the lead-based reactor under different operating conditions; S106, Based on the core geometry model containing bubbles, the reaction cross section is calculated; S108, homogenize the reaction cross section to generate the reaction cross section data.

[0010] Optionally, step S120, which involves calculating the core power distribution based on the reaction cross-section data and a core neutronics calculation model, includes: S122, Fit the reaction cross-section data to obtain a continuous function of the reaction cross-section with respect to the cavitation fraction; S124, based on the continuous function and the core neutron flux density, calculate the core neutron flux density; and based on the core neutron flux density, calculate the core power distribution.

[0011] Optionally, step S122, fitting the reaction cross-section data to obtain a continuous function of the reaction cross-section with respect to the cavitation fraction, includes: S123, perform polynomial fitting on the reaction cross-section data to obtain a continuous function of the reaction cross-section with respect to the cavitation fraction; the continuous function is: ; In the formula, For the reaction cross section, This refers to the fraction of voids; These are the fitting coefficients. .

[0012] Optionally, after step S140, which involves obtaining the thermal parameters for the current time step based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model, the method further includes: S160, determine whether the core power distribution, the fuel rod temperature, the coolant temperature and the void fraction have converged. If so, proceed to the next time step and repeat steps S120 and S140.

[0013] Optionally, between step S140 and step S160, the method further includes: S150: Determine whether the current time step has reached the preset time. If yes, output the result; otherwise, execute step S160.

[0014] This invention provides a method for calculating the gas-liquid two-phase nuclear-thermal coupling of lead-based reactor cores, which has at least the following beneficial technical effects: By generating reaction cross-section data with multiple groups, multiple temperature points, and multiple void fractions, and transferring this data to the core neutronics calculation model, the core power distribution is calculated. This core power distribution is then transferred to the core gas-liquid two-phase thermal-hydraulic calculation model to obtain thermal parameters. This calculation method can accurately simulate the influence of gas phase on neutronics characteristics, and thus accurately predict the core power distribution and thermal parameters under gas-liquid two-phase flow conditions. This addresses the power transient safety issues that may arise in the core under severe accidents caused by gas-liquid two-phase flow, providing technical support for the safety analysis and accident assessment of lead-based reactors.

[0015] A second aspect of the present invention provides a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation system, which applies the above-described lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method, the system comprising: The geometry model building module is used to build geometry models; The cross-section data generation module is used to generate reaction cross-section data based on the core geometry model. The core power distribution calculation module is used to calculate the core power distribution based on the reaction cross-section data and a core neutronics calculation model; and, The thermal parameter calculation module obtains the thermal parameters for the current time step based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model. The thermal parameters include fuel rod temperature, coolant temperature, and void fraction.

[0016] This invention provides a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation system, which applies the lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method described above and has all the advantages of the above coupling calculation method, which will not be repeated here. Attached Figure Description

[0017] Figure 1 A flowchart illustrating a lead-based reactor core gas-liquid two-phase nuclear-thermal coupling calculation method provided in this embodiment of the invention. Figure 1 ; Figure 2 A flowchart illustrating the steps of constructing a core geometric model and generating reaction section data in a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method provided in an embodiment of the present invention. Figure 3 A flowchart illustrating the steps of calculating the core power distribution based on the reaction cross-section data and the core neutronics calculation model in a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method provided in an embodiment of the present invention. Figure 4 In a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method provided in an embodiment of the present invention, a flowchart is shown in which the thermal parameters of the current time step are obtained based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model. Figure 5 A flowchart illustrating a lead-based reactor core gas-liquid two-phase nuclear-thermal coupling calculation method provided in this embodiment of the invention. Figure 2 ; Figure 6 The diagram shows the change of power distribution in a lead-based reactor core over time in a gas-liquid two-phase nuclear thermal coupling calculation method provided in this embodiment of the invention. Figure 7 This is a block diagram of a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation system provided in an embodiment of the present invention.

[0018] Explanation of reference numerals in the attached figures: 10. Geometric model construction module; 20. Cross-section data generation module; 30. Core power distribution calculation module; 40. Thermal parameter calculation module. Detailed Implementation

[0019] To make the above-mentioned objects, features and advantages of the present invention more apparent and understandable, the following description is provided in conjunction with the accompanying drawings. Figures 1-7 Specific embodiments of the present invention will be described in detail below.

[0020] This invention provides a method for calculating the gas-liquid two-phase nuclear-thermal coupling of a lead-based reactor core, see appendix. Figure 1 The method includes: S100, Construct the core geometry model and generate reaction cross-section data. In this step, the core geometry model includes the fuel rod region and the coolant region, and the reaction cross-section data includes multiple groups, multiple temperature points, and multiple void fractions. This step generates the reaction cross-section data based on OpenMC, an open-source Monte Carlo particle transport simulation software primarily used in nuclear reactor physics that supports geometric modeling. For details, see the appendix. Figure 2 The specific implementation process of step S100 is as follows: Step S102: Construct the core geometry model; specifically, based on the Python API of OpenMC, construct a core geometry model consistent with the actual core structure; the core geometry model includes the fuel rod region and the coolant region; it should be noted that the core geometry model is a two-dimensional core geometry model, and the cross-section of the core geometry model is a regular hexagon. Step S104: Insert bubbles into the coolant region to obtain a core geometry model containing bubbles; this setting is to simulate the gas-liquid two-phase flow state of the lead-based reactor under different operating conditions. Step S106: Calculate the reaction cross section based on the core geometry model containing bubbles; Step S108 involves homogenizing the reaction cross section to generate reaction cross section data, including multi-group, multi-temperature point, and multi-cavitation fraction data. Specifically, the reaction cross section is subjected to SPH correction to homogenize it, resulting in multi-group, multi-temperature point, and multi-cavitation fraction reaction cross section data. This setup yields a core geometry model containing bubbles and generates multi-group, multi-temperature point, and multi-cavitation fraction reaction cross section data, providing input for subsequent neutronics calculations. This allows for accurate simulation of the influence of gas phases on neutronics properties, and precise prediction of core power distribution and thermal parameters under gas-liquid two-phase flow conditions, providing technical support for the safety analysis and accident assessment of lead-based reactors.

[0021] S120, Based on the reaction cross-section data and the core neutronics calculation model, calculate the core power distribution; in this step, the reaction cross-section data is transferred to the Morphy neutronics calculation model to calculate the core power distribution; for details, see Appendix. Figure 3 The specific implementation process of step S120 is as follows: Step S122 involves fitting the reaction cross-section data to obtain a continuous function of the reaction cross-section with respect to the cavitation fraction. It should be noted that the fitting method can be nonlinear fitting, polynomial fitting, or other existing fitting methods. For example, step S122 could be: S123, performing polynomial fitting on the reaction cross-section data to obtain a continuous function of the reaction cross-section with respect to the cavitation fraction; this continuous function is: ; In the formula, For the reaction cross section, This refers to the fraction of voids; These are the fitting coefficients. .

[0022] Step S124: Based on the aforementioned continuous function and the core neutron flux calculation model, calculate the core neutron flux density; and based on the core neutron flux density, calculate the core power distribution. It should be noted that this core neutron flux calculation model uses a neutron flux calculation model from the prior art, which will not be elaborated upon here.

[0023] S140, based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model, the thermal parameters of the current time step are obtained, including fuel rod temperature, coolant temperature and void fraction.

[0024] The lead-based reactor core gas-liquid two-phase nuclear-thermal coupling calculation method provided in this invention generates reaction cross-section data with multiple groups, multiple temperature points, and multiple vacuolar fractions. This reaction cross-section data is then transferred to the core neutronics calculation model to calculate the core power distribution. This core power distribution is then transferred to the core gas-liquid two-phase thermal-hydraulic calculation model to obtain thermal parameters. This calculation method can accurately simulate the influence of gas phase on neutronics characteristics, and thus accurately predict the core power distribution and thermal parameters under gas-liquid two-phase flow conditions. This addresses the power transient safety issues that may arise from gas-liquid two-phase flow in the core under severe accidents, providing technical support for the safety analysis and accident assessment of lead-based reactors.

[0025] In this embodiment of the invention, in step S140, the core gas-liquid two-phase thermal-hydraulic calculation model includes a thermal-hydraulic model and a transport model. The thermal-hydraulic model is used to calculate the fuel rod temperature and coolant temperature. This thermal-hydraulic model adopts the thermal-hydraulic model in the prior art, which will not be described in detail here. The transport model includes the gas-liquid two-phase mass conservation equation, the gas-liquid two-phase momentum conservation equation, and the gas-liquid two-phase energy conservation equation, which are used to calculate the void fraction. It should be noted that the transport model is constructed based on the simplified assumptions of "no slip ratio in the gas-liquid two-phase system" and "no interphase heat transfer".

[0026] Specifically, since the velocities of the gas and liquid phases are equal, and the velocities of the gas phase and liquid phase are equal, the mass conservation equation for the gas and liquid phases is: ; In the formula, Let be the density of the gas-liquid two-phase mixture. For time, The velocity of the gas and liquid phases, For the fraction of voids, For gas phase density, The density of the liquid phase; Since there is no slip ratio between the gas and liquid phases, there is no relative motion between them. Therefore, the effects of interphase drag and lift can be ignored. Thus, the momentum conservation equation for the gas-liquid two-phase system is: ; In the formula, Let be the density of the gas-liquid two-phase mixture. For time, The velocity of the gas and liquid phases, For flow field pressure, The dynamic viscosity of the gas-liquid two-phase mixture is denoted as . For the fraction of voids, The dynamic viscosity of the gas phase. The dynamic viscosity of the liquid phase. It is the acceleration due to gravity; Since there is "no interphase heat transfer" and "no slip ratio" between the gas and liquid phases, and their spatial distributions are synchronized, the energy conservation equation for the gas-liquid two-phase system can be simplified to: ; In the formula, Let be the density of the gas-liquid two-phase mixture. Enthalpy of a gas-liquid two-phase mixture For time, The velocity of the gas and liquid phases, Enthalpy of a gas-liquid two-phase mixture For the fraction of voids, For vapor phase enthalpy, For liquid phase enthalpy, Thermal conductivity, The characteristic temperature of the coolant. This is the rate at which heat is released from fuel to coolant per unit volume.

[0027] The lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method provided in this embodiment of the invention can achieve refined simulation of gas-liquid two-phase flow and heat transfer in lead-based reactor cores by developing a deep fusion of transport model and thermal-hydraulic model.

[0028] In this embodiment of the invention, see appendix. Figure 4 Step S140, the step of obtaining the thermal parameters for the current time step based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model, includes: S142, based on the core power distribution and the thermal-hydraulic model, the fuel rod temperature and coolant temperature are calculated. S144, solve the gas-liquid two-phase mass conservation equation, gas-liquid two-phase momentum conservation equation, and gas-liquid two-phase energy conservation equation in sequence to calculate the cavitation fraction.

[0029] In this embodiment of the invention, after step S140, which involves obtaining the thermal parameters for the current time step based on the core power distribution and the core gas-liquid two-phase thermo-hydraulic calculation model, please refer to the appendix. Figure 5 The methods also include: S160: Determine whether the core power distribution, fuel rod temperature, coolant temperature, and void fraction have converged. That is, whether the change in core power distribution in adjacent time steps meets the preset power distribution difference, the change in fuel rod temperature in adjacent time steps meets the preset fuel rod temperature difference, the change in coolant temperature in adjacent time steps meets the preset coolant temperature difference, and the change in void fraction in adjacent time steps meets the preset void fraction difference. If yes, proceed to the next time step and repeat steps S120 and S140. If no, modify the reaction cross section data and recalculate the core power distribution based on the modified reaction cross section data and the core neutronics calculation model.

[0030] This setup is repeated iteratively until the changes in core power distribution, fuel rod temperature, and coolant temperature meet the convergence conditions, thereby achieving a precise coupled simulation of the physical processes and thermal-hydraulic processes of the core under gas-liquid two-phase conditions.

[0031] In this embodiment of the invention, see appendix. Figure 4 Between step S140 and step S160, the method further includes: S150: Determine whether the current time step has reached the preset time. If yes, output the result; otherwise, execute step S160.

[0032] To further illustrate the present invention, the calculation method for gas-liquid two-phase nuclear thermal coupling of lead-based reactor core provided by the present invention will be described in more detail below with reference to the accompanying drawings and embodiments, but these should not be construed as limiting the scope of protection of the present invention.

[0033] Appendix Figure 6 This diagram illustrates the change in power distribution over time in a lead-based reactor core using a gas-liquid two-phase nuclear-thermal coupling calculation method provided in an embodiment of the invention. When bubbles enter the core, the core power distribution rapidly increases due to the increased fission cross-section and decreased absorption cross-section, reaching a maximum power of 180% of the original power. Later, the negative reactivity coefficient, resulting from the increased temperatures of both fuel rods and coolant, begins to take effect, and the power gradually decreases. The power distribution calculation results based on the gas-liquid two-phase nuclear-thermal coupling calculation method for lead-based reactor cores provided in this embodiment of the invention are consistent with expectations, namely, the power surge phenomenon caused by bubbles entering the core after a lead-based reactor accident.

[0034] This invention also provides a lead-based reactor core gas-liquid two-phase nuclear-thermal coupling calculation system, which applies the above-mentioned lead-based reactor core gas-liquid two-phase nuclear-thermal coupling calculation method, see appendix. Figure 7The system includes a geometric model construction module 10, a cross-section data generation module 20, a core power distribution calculation module 30, and a thermal parameter calculation module 40. The geometric model construction module 10 is used to construct the geometric model; the cross-section data generation module 20 is used to generate reaction cross-section data based on the core geometric model; the core power distribution calculation module 30 is used to calculate the core power distribution based on the reaction cross-section data and the core neutronics calculation model; and the thermal parameter calculation module 40 is used to obtain the thermal parameters for the current time step based on the core power distribution and the core gas-liquid two-phase thermal-hydraulic calculation model. The thermal parameters include fuel rod temperature, coolant temperature, and void fraction.

[0035] The lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation system provided in this embodiment of the invention applies the above-mentioned lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method and has all the beneficial effects of the calculation method, which will not be elaborated here.

[0036] While the present invention has been disclosed above, it is not limited thereto. Any person skilled in the art can make various modifications and alterations without departing from the spirit and scope of the invention; therefore, the scope of protection of the present invention should be determined by the scope defined in the claims.

Claims

1. A method for calculating gas-liquid two-phase nuclear heat coupling in a lead-based reactor core, characterized in that, The method comprises: S100, constructing a core geometry model to generate reaction cross-section data; wherein the core geometry model comprises a fuel rod region and a coolant region, and the reaction cross-section data comprises multiple groups, multiple temperature points, and multiple void fractions; S120, calculating a core power distribution based on a core neutronics calculation model according to the reaction cross-section data; S140, obtaining thermal parameters of a current time step based on a core gas-liquid two-phase thermal-hydraulic calculation model according to the core power distribution, wherein the thermal parameters comprise a fuel rod temperature, a coolant temperature, and a void fraction; the core gas-liquid two-phase thermal-hydraulic calculation model comprises a thermal-hydraulic model and a transport model, the thermal-hydraulic model is used to calculate the fuel rod temperature and the coolant temperature, and the transport model comprises a gas-liquid two-phase mass conservation equation, a gas-liquid two-phase momentum conservation equation, and a gas-liquid two-phase energy conservation equation, and is used to calculate the void fraction; wherein the gas-liquid two-phase mass conservation equation is: ； wherein is the density of the gas-liquid two-phase mixture, is time, is the velocity of the gas-liquid two-phase, is the void fraction, is the gas phase density, is the liquid phase density; the gas-liquid two-phase momentum conservation equation is: ； wherein is the density of the gas-liquid mixture, is time, is the velocity of the gas-liquid mixture, is the pressure of the flow field, is the dynamic viscosity of the gas-liquid mixture, is the void fraction, is the dynamic viscosity of the gas phase, is the dynamic viscosity of the liquid phase, is the gravitational acceleration; the gas-liquid two-phase energy conservation equation is: ； wherein is the density of the gas-liquid two-phase mixture, is the enthalpy of the gas-liquid two-phase mixture, is time, is the velocity of the gas-liquid two-phase, is the enthalpy of the gas-liquid two-phase mixture, is the void fraction, is the enthalpy of the gas phase, is the enthalpy of the liquid phase, thermal conductivity, is the coolant characteristic temperature, is the heat release rate of the fuel to the coolant per unit volume.

2. The lead-based reactor core gas-liquid two-phase thermal coupling calculation method according to claim 1, characterized in that, The step S140 of obtaining the thermal parameters of the current time step based on the core gas-liquid two-phase thermal-hydraulic calculation model according to the core power distribution comprises: S142, calculating the fuel rod temperature and the coolant temperature based on the thermal-hydraulic model according to the core power distribution; S144, sequentially solving the gas-liquid two-phase mass conservation equation, the gas-liquid two-phase momentum conservation equation, and the gas-liquid two-phase energy conservation equation to calculate the void fraction.

3. The lead-based reactor core gas-liquid two-phase flow heat coupling calculation method according to claim 1, characterized in that, The step S100 of constructing the core geometry model to generate the reaction cross-section data comprises: S102, constructing the core geometry model; S104, inserting a gas bubble into the coolant region to obtain a core geometry model containing a gas bubble; S106, calculating a reaction cross-section based on the core geometry model containing the gas bubble; S108, performing homogenization processing on the reaction cross-section to generate the reaction cross-section data.

4. The lead-based reactor core gas-liquid two-phase thermal coupling calculation method according to any one of claims 1-3, characterized in that, The step S120 of calculating the core power distribution based on the core neutronics calculation model according to the reaction cross-section data comprises: S122, fitting the reaction cross-section data to obtain a continuous function of the reaction cross-section with respect to the void fraction; S124, calculating a core neutron flux density based on the core neutronics calculation model according to the continuous function, and calculating the core power distribution based on the core neutron flux density.

5. The lead-based reactor core gas-liquid two-phase flow heat coupling calculation method according to claim 4, characterized in that, The step S122 of fitting the reaction cross-section data to obtain the continuous function of the reaction cross-section with respect to the void fraction comprises: S123, performing polynomial fitting on the reaction cross-section data to obtain a continuous function of the reaction cross-section with respect to the void fraction; and the continuous function is: ； wherein is the reaction cross section, is the void fraction; is the fitting coefficient, .

6. The lead-based reactor core gas-liquid two-phase thermal coupling calculation method according to any one of claims 1-3, characterized in that, After the step S140 of obtaining the thermal parameters of the current time step based on the core gas-liquid two-phase thermal-hydraulic calculation model according to the core power distribution, the method further comprises: S160, judging whether the core power distribution, the fuel rod temperature, the coolant temperature and the void fraction converge, if yes, performing next time step, repeating steps S120 and S140.

7. The lead-based reactor core gas-liquid two-phase flow heat coupling calculation method according to claim 6, characterized in that, Between the step S140 and the step S160, the method further comprises: S150, judging whether the current time step reaches a preset time, if yes, outputting a result; otherwise, performing step S160.

8. A lead-based reactor core gas-liquid two-phase nuclear heat coupling calculation system, characterized in that, The system comprises: a geometry model constructing module (10) configured to construct a geometry model; a cross section data generating module (20) configured to generate reaction cross section data based on the core geometry model; a core power distribution calculating module (30) configured to calculate a core power distribution based on a core neutronics calculation model according to the reaction cross section data; and a thermal parameter calculating module (40) configured to obtain thermal parameters of a current time step based on a core thermal-hydraulic calculation model of gas-liquid two-phase according to the core power distribution, wherein the thermal parameters comprise a fuel rod temperature, a coolant temperature and a void fraction.

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