Multi-physics field coupling system and method for micro reactor

Abstract

A multi-physics coupling system and method for a microreactor includes: a neutron physics simulation module, a thermal-hydraulic simulation module, a power conversion module, a main function module, a burnup nuclide update module, and a control drum position update module. The present invention dynamically adjusts the position of the control drum via the control drum position update module during the burnup coupling calculation process. In conventional multi-physics coupling methods, the position of the control drum is fixed, so the effective proliferation factor of the core is always greater than 1.0, and the results of the coupling calculation cannot reflect the actual state of the reactor during operation. When the multi-physics coupling method of the present invention is used for calculation, by continuously adjusting the control drum position, the effective proliferation factor of the reactor is always near 1.0, which is closer to the operating conditions of a real reactor and can more accurately simulate the state of the reactor.

Description

Technical Field

[0001] The present invention relates to a technology in the field of micro reactor control, in particular to a multi-physical field coupling system and method for micro reactors. Background Art

[0002] Existing multi-physics coupling methods have never taken reactor burnup into account, because the power distribution and nuclides at the beginning, middle, and end of a reactor's life are quite different. Microreactors usually control reactivity through a control drum. As the reactor continues to operate, the reactivity gradually decreases. At this time, the position of the control drum needs to be adjusted to keep the reactor in a critical state. However, conventional coupling does not take into account the adjustment of the control drum position. Summary of the Invention

[0003] In view of the fact that the existing technology cannot adjust the position of the control drum according to the fuel consumption and the critical state of the reactor, the present invention proposes a multi-physics field coupling system and method for a micro reactor. On the basis of the conventional coupling method, a fuel consumption nuclide update module and a control drum position update module are added, which can dynamically couple the fuel consumption and adjust the position of the control drum according to the reactivity.

[0004] The present invention is achieved through the following technical solutions:

[0005] The present invention relates to a multi-physics field coupling system for a micro reactor, comprising: a neutron physics simulation module, a thermal hydraulic simulation module, a power conversion module, a main function module, a burnup nuclide update module, and a control drum position update module, wherein: the neutron physics simulation module receives the fuel temperature field, coolant temperature field, and coolant density field data output by the thermal hydraulic program and converts them into a data format recognizable by the neutron physics simulation module, generates and updates the neutron physics calculation model, calls and runs the neutron physics calculation program, and obtains the fission power shape distribution of the core, the nuclide information of the burnup, and the critical state information of the core; the power conversion module receives The fission power distribution generated by the neutron physics simulation is converted into a data format that can be recognized and applied by the thermal-hydraulic program, and then passed to the thermal-hydraulic simulation module. After receiving the data, the thermal-hydraulic simulation module calculates and outputs the fuel temperature field, coolant temperature field and coolant density field. The burnup nuclide update module converts the data according to the nuclide information generated by the neutron physics program and adds it back into the neutron physics calculation model. The control drum position update module obtains the core critical state information according to the neutron physics calculation program, solves the control drum value function, and updates the control drum position in the neutron physics calculation model according to the coordinates and diameter of the control drum.

[0006] The present invention relates to a multi-physics field coupling method based on the above system, comprising the following steps:

[0007] Step 1) Based on the neutron physics program, the neutron macroscopic cross sections of the fuel and coolant are obtained according to the initial fuel temperature, fuel density, coolant temperature, and coolant density. Then, the neutron transport equation and the burnup equation are solved to calculate the neutron fission power, the effective proliferation factor of the core, and the distribution and ratio of the burnup nuclides.

[0008] Step 2) Convert the neutron fission rate into power distribution, and then obtain the power of each grid.

[0009] Step 3) Assign the power of each grid to the thermal grid, and obtain the fuel temperature, fuel density, coolant temperature, and coolant density by solving the mass conservation equation, momentum conservation equation, and energy conservation equation;

[0010] Step 4) When the temperature field and density field of the fuel and coolant obtained in step 3) converge, proceed to step 5); otherwise, update the fuel temperature, coolant temperature, and coolant density in the neutron physics model file and return to step 1);

[0011] Step 5) When the preset burnup is reached, the calculation ends, and the effective growth factor, power distribution, fuel temperature, coolant temperature, coolant density, and control drum position change information for the entire reactor burnup life are obtained. If the preset number of burnup steps has not been reached, step 6) is executed to continue the calculation.

[0012] Step 6) Run the burnup update module and the control drum position update module, update the fuel temperature, coolant temperature, and coolant density in the neutron physics model file, and proceed to the next step of burnup calculation;

[0013] Technical Effects

[0014] The present invention updates the position of the control drum in real time based on fuel consumption and the critical state of the core during multi-physics coupling calculations of neutron physics and thermal hydraulics. Existing multi-physics coupling methods do not consider position adjustments of the control drum or control rods when performing fuel consumption coupling calculations, so the calculated effective proliferation factor of the core shows a generally linearly decreasing trend. During the coupling calculation process, the effective proliferation factor of the core is always greater than 1, meaning the core is always in a supercritical state. The method of the present invention adjusts the position of the control drum based on fuel consumption and the critical state of the core. Under ideal conditions, the effective proliferation factor of the core can be maintained near 1, which is closer to the actual state of reactor operation. During the coupling process, the angle of the control drum is continuously adjusted, from an initial 113.2° to a final 168.3°, which is not available in conventional multi-physics coupling methods. BRIEF DESCRIPTION OF THE DRAWINGS

[0015] Figure 1 Flowchart of the present invention;

[0016] Figure 2Schematic diagram of the multi-physics coupling system for micro reactors;

[0017] Figure 3 is a schematic diagram of a microreactor according to an embodiment;

[0018] Figure 4 This is a schematic diagram of the control drum;

[0019] In the figure: reflective layer 1, control drum unit 2, cladding 3, lead-bismuth coolant 4, dispersed fuel 5, shutdown control rod 6, control drum structural material 7, control drum neutron absorbing material coating 8. DETAILED DESCRIPTION

[0020] like Figure 2 As shown, a multi-physics field coupling system for a micro reactor involved in this embodiment includes: a neutron physics simulation module, a thermal hydraulic simulation module, a power conversion module, a main function module, a burnup nuclide update module, and a control drum position update module, wherein: the neutron physics simulation module receives the fuel temperature field, coolant temperature field, and coolant density field data output by the thermal hydraulic program and converts them into a data format recognizable by the neutron physics simulation module, generates and updates the neutron physics calculation model, calls and runs the neutron physics calculation program to obtain the core fission power shape distribution, the burnup nuclide information, and the core critical state information; the power conversion module The fission power distribution generated by the neutron physics simulation is received and converted into a data format that can be recognized and applied by the thermal-hydraulic program, and then passed to the thermal-hydraulic simulation module. After receiving the data, the thermal-hydraulic simulation module calculates and outputs the fuel temperature field, coolant temperature field and coolant density field. The burnup nuclide update module converts the data according to the nuclide information generated by the neutron physics program and adds it back into the neutron physics calculation model. The control drum position update module obtains the core critical state information according to the neutron physics calculation program, solves the control drum value function, and updates the control drum position in the neutron physics calculation model according to the coordinates and diameter of the control drum.

[0021] This embodiment develops a multi-physics coupling program for neutron physics and thermal hydraulics based on programming languages (including but not limited to Matlab, Python, Fortran, and C++). Based on this coupling program, a fuel consumption coupling module, a fuel consumption nuclide update module, and a control drum position update module are developed. Due to differences in underlying design languages, application scenarios, and developer habits between the neutron physics and thermal hydraulics programs, the modeling methods and data formats of the two programs differ significantly. Furthermore, the power distribution of the neutron physics program is significantly affected by the coolant temperature, density, and fuel temperature calculated by the thermal hydraulics program. Furthermore, the coolant temperature, density, and fuel temperature are in turn affected by the power distribution. The two programs interact with each other, resulting in a strong coupling effect. To improve simulation accuracy, a multi-physics coupling program is necessary. The coupling program processes the data generated by the neutron physics and thermal hydraulics programs, converting the data into a mutually recognizable and usable format; calls the neutron physics and thermal hydraulics programs; determines whether the calculations have converged and terminated; and calls the fuel consumption nuclide update module and the control drum position update module to update the neutron physics calculation model. The burnup nuclide update module is also developed based on a programming language. It converts the format of the nuclide file generated by neutron physics and inputs it into the neutron physics calculation model. The control drum position update module adjusts the position of the control drum in real time according to the critical state of the reactor.

[0022] like Figure 1 As shown, this embodiment relates to a multi-physics field coupling method for a micro reactor based on the above system, comprising the following steps:

[0023] Step 1) Based on the neutron physics simulation module, according to the initial fuel temperature Fuel density Coolant temperature Coolant density By evaluating the nuclear database, the neutron macroscopic cross section of the fuel and coolant is obtained, a neutron physics model file is generated, and the neutron physics program is called to solve the neutron transport equation and the burnup equation to calculate the fission power. and burnup nuclide files Where: p is the p-th burnup, q is the number of iterations within the p-th burnup step, and g is the grid number.

[0024] The neutron physics programs include but are not limited to: OpenMC, MCNP, RMC, Serpent and other Monte Carlo programs.

[0025] The neutron physics model file refers to an input card that can be received by the neutron physics program, which defines the structure, size, material, and boundary conditions of the calculation model.

[0026] The evaluation nuclear database includes but is not limited to: China's CENDL and the United States' ENDF database.

[0027] The neutron transport equation is: in: For the location The energy is E and the direction of motion is The neutron angular fluence rate, For the location The macroscopic cross section of neutrons with energy E is determined by the fuel temperature. Fuel density Coolant temperature Coolant density By evaluating the nuclear database, For the location The energy is E and the direction of motion is A scattered neutron source, For the location The energy is E and the direction of motion is When solving the neutron transport equation, given the neutron macroscopic cross section and the fission neutron source, the neutron fluence rate can be calculated. The neutron fluence rate multiplied by the neutron fission cross section is Fission power can be obtained

[0028] The fuel consumption equation is: Where: N i (t) is the concentration of nuclide i at time t, l ij is the proportion of nuclide i formed by radioactive decay in nuclide j, λ j is the decay constant of nuclide j, N j is the concentration of nuclide j, f ik is the proportion of nuclide i formed by neutron reaction in nuclide k, φ is the average neutron fluence rate, σ k is the microscopic cross section of nuclide k, N k is the concentration of nuclide k, λ i is the decay constant of nuclide i, σ i is the microscopic cross section of nuclide i, N i represents the concentration of nuclide i. In the burnup equation, M represents the number of nuclides of interest. A typical reactor produces over 1,500 nuclides, and the number of nuclides of interest can be adjusted based on the computing hardware.

[0029] Step 2) Convert the fission power to Convert to power distribution Since the fission power obtained by neutron physics program statistics is only a shape distribution, it needs to be converted into a power distribution. The sum of the fission power and the sum of the power are linearly related, and the coefficient S is obtained, and then That is the power of each grid.

[0030] The power conversion module specifically extracts the fission power of each grid in the Tally format file calculated by the neutron physics program, and then converts it into the power of each grid through step 2) to form a format file that can be read by the thermal hydraulic simulation module.

[0031] Step 3) Use the thermal hydraulic simulation module to calculate the power of each grid Assign the thermal grid, and through the thermal hydraulic simulation module, load the power of each grid calculated by the power conversion module into the model file of the thermal hydraulic program, call the thermal hydraulic program, solve the mass conservation equation, momentum conservation equation and energy conservation equation, and obtain the fuel temperature Fuel density Coolant temperature and coolant density

[0032] The thermal hydraulic programs include but are not limited to: Fluent, STAR-CCM+.

[0033] The mass conservation equation is:

[0034] The momentum conservation equation is:

[0035]

[0036] The energy conservation equation is: in: are the velocities in the x, y, and z directions, μ is the viscosity, and f x , f y , f z is the external force on the grid, T is the grid temperature, λ is the thermal conductivity of the fluid, c p is the constant pressure specific heat capacity of the fluid, S T is the viscous dissipation term.

[0037] Step 4) When the temperature field and density field of the fuel and coolant obtained in step 3) converge, proceed to step 5); otherwise, set the fuel temperature to Fuel density Coolant temperature Coolant density Import the neutron physics simulation module and return to step 1);

[0038] The convergence mentioned here refers to: in: are the residuals of fuel temperature, coolant temperature, and coolant density, respectively. Calculate the maximum residuals allowed for fuel temperature, coolant temperature, and coolant density respectively.

[0039] Step 5) When the burnup time reaches a preset value, the calculation ends and the effective proliferation factor, power distribution, fuel temperature, coolant temperature, coolant density, and control drum position change information of the reactor throughout the burnup life are obtained. Otherwise, step 6) is executed.

[0040] The burnup time is the time the reactor operates at full power, in days.

[0041] The effective multiplication factor (EPF) is the ratio of the number of neutrons in the new generation of the reactor core to the number of neutrons in the immediately preceding generation. When EPF is greater than 1.0, the reactor is in a supercritical state and power is increasing. When EPF is 1.0, power remains constant and the reactor is in a stable state. When EPF is less than 1.0, power is decreasing. In actual reactor operation, EPF is 1.0.

[0042] Step 6) The burnup nuclide update module updates the neutron physics model file through the neutron physics simulation module based on the burnup nuclide file generated in step 1), and calls the neutron physics program to perform the next burnup calculation, specifically: according to the fuel temperature obtained in step 4) Fuel density Coolant temperature and coolant density And the nuclide data obtained in step 6), and call the control drum position update module to adjust the angle of the control drum, run the neutron physics simulation module, generate a new neutron physics model file, run the neutron physics program, perform the p+1 step burnup calculation, and proceed to step 1).

[0043] Data conversion refers to converting nuclide information into nuclide data. For example, in the example nuclide information generated by a neutron physics program, the first line contains the grid number (cell=1>9>45123), the material number (mat=22593), and the second line contains the nuclide code (NuclideMassDensityAtomDensity). The first two digits of the nuclide code represent the number of protons, the third to fourth digits (if the mass number is greater than 100, the third to fifth digits) represent the mass number, and the last digit is invalid information. For example, 420960 represents a nuclide with 42 protons and a mass number of 96, which is molybdenum.

[0044] Table 1 Nuclide information generated by neutron physics program

[0045] cell=1>9>45123 mat=22593 Nuclide MassDensity AtomDensity Sum 1.16379E+01 6.31320E-02 420960 4.62722E+00 2.90552E-02 220480 2.32758E-02 2.92210E-04 400910 3.72412E-03 2.46704E-05 922350 5.68124E+00 1.45559E-02 922380 9.09524E-01 2.30085E-03 70140 3.91964E-01 1.68565E-02

[0046] Table 2 Nuclide data processed by the nuclide update module

[0047] mat=22593 -1.16379E+01 42096.72c -4.62722E+00 22048.72c -2.32758E-02 40091.72c -3.72412E-03 92235.72c -5.68124E+00 92238.72c -9.09524E-01 7014.72c -3.91964E-01

[0048] The first line of information after conversion indicates that the density of material No. 45123 is 11.6375 g / cm 3 , which includes nuclides such as 42096 and 22048. The suffix .72c represents the nuclide cross section at 900K. The mass densities of the nuclides are 4.6272 and 0.023276, respectively. The minus sign (-) is required for the density and mass density in the above figures because neutron physics programs distinguish between mass density and atomic density. The negative sign indicates mass density, and the positive sign indicates atomic density.

[0049] The control drum value function means that when the control drum faces the core at different angles, its impact on the core is different. This impact can be expressed by a polynomial, namely the control drum value curve. By solving the polynomial, the angle of the control drum is obtained.

[0050] like Figure 3 As shown, the microreactor involved in this embodiment includes: a reflector 1 and a hexagonal fuel 5 arranged therein, six control drum units 2 for reactivity control located outside the fuel 5, and a plurality of lead-bismuth coolants 4 located inside the fuel 5, and a control rod 6 for emergency shutdown, wherein: the control drum units 2 can be divided into two groups, 1, 3, 5 as one group, and 2, 4, 6 as another group, and the two groups of control drum units are rotated by independent drive mechanisms.

[0051] like Figure 3 and Figure 4As shown, the control drum unit 2 includes: a main body material 8 and a neutron absorbing material 9 located outside the main body material 8. The control drum unit is obtained by Boolean operation of two circles and two straight lines, as shown in FIG. Figure 4 As shown. The control drum in the figure is as follows in the neutron physics model file: -R_1∩+R_2∩-L_1∩-L_2, the material is B4C, ! (-R_1∩+R_2∩-L_1∩-L_2), the material is BeO, where: the negative sign (-) indicates the inside of the circle or below the line, the positive sign (+) indicates the outside of the circle or above the line, and ! indicates negation. Therefore, -R_1∩+R_2∩-L_1∩-L_2 is expressed as the inside of circle R_1, the outside of circle R_2, the bottom of line L_1, and the bottom of line L_2 intersect to obtain Figure 4 The black part in the graph is made of B4C! (-R_1∩+R_2∩-L_1∩-L_2) can be expressed as the material of the part other than the black part, which is BeO.

[0052] When the center coordinates of control drum unit 1 are (0, D), the arc angle of the neutron absorber is θ, the angle of the control drum facing the core is β, the radius of circle R_1 is R, and the radius of circle R_2 is r, then:

[0053] The control equations of the control drum unit No. 1 include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: 2 +(yD) 2 =R 2 ; The equation of circle R_2 is: 2 +(yD) 2 =r 2 The above four equations can be used to obtain the control drum facing the core at any angle.

[0054] The control equations of the control drum unit No. 2 include: The equation of the straight line L_1 is:

[0055] The equation of line L_2 is: The equation of circle R_1 is: The equation of circle R_2 is:

[0056] The control equations of the control drum unit No. 3 include: The equation of the straight line L_1 is: The equation of line L_2 is:

[0057] The equation of circle R_1 is: The equation of circle R_2 is:

[0058] The control equations of the No. 4 control drum unit include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: 2 +(y+D) 2 =R 2 ; The equation of circle R_2 is: 2 +(y+D) 2 =r 2 ;

[0059] The control equations of the No. 5 control drum unit include: The equation of the straight line L_1 is:

[0060] The equation of line L_2 is: The equation of circle R_1 is: The equation of circle R_2 is:

[0061] The control equations of the No. 6 control drum unit include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: The equation of circle R_2 is:

[0062] When the six control drum units have R values of 8.0 cm, r values of 7.0 cm, and D values of 19.75 cm, the neutron absorber arc angle is 120°, and the burnup duration is 1800 days, coupled calculations using the method of the present invention yielded the results shown in Tables 3 and 4. During the burnup coupled calculations, due to the continuous adjustment of the control drum angles, the effective growth factor of the core remained near 1.0, closer to the actual operating conditions of the reactor. Adjustments by the control drum position update module increased the control drum angle from an initial 113.2° to a final 168.3°.

[0063] Table 3 Effects of conventional multi-physics coupling methods

[0064] Burn-up days Core effective growth factor Control drum angle (°) 365 1.012907 180 730 1.010018 180 1095 1.007105 180 1460 1.003989 180 1825 1.000855 180

[0065] Table 4 Effects of the multi-physics field coupling method of the present invention

[0066]

[0067]

[0068] Compared to existing technologies, this method dynamically adjusts the control drum position via a control drum position update module during the burnup coupling calculation process. Conventional multiphysics coupling methods, however, keep the control drum position fixed, resulting in a core effective proliferation factor consistently exceeding 1.0. These coupled calculations fail to reflect the actual operating state of the reactor. However, using the multiphysics coupling method of the present invention, by continuously adjusting the control drum position, the reactor's effective proliferation factor remains near 1.0, more closely resembling actual reactor operating conditions and enabling more accurate simulation of the reactor's state.

[0069] The above-mentioned specific implementation can be partially adjusted in different ways by those skilled in the art without departing from the principles and purpose of the present invention. The scope of protection of the present invention shall be based on the claims and shall not be limited by the above-mentioned specific implementation. All implementation schemes within its scope shall be subject to the constraints of the present invention.

Claims

1. A multi-physics field coupling system for a microreactor, characterized in that: include: Neutron physics simulation module, thermal hydraulic simulation module, power conversion module, main function module, fuel consumption nuclide update module and control drum position update module, wherein: the neutron physics simulation module receives the fuel temperature field, coolant temperature field and coolant density field data output by the thermal hydraulic program and converts them into a data format recognizable by the neutron physics simulation module, generates and updates the neutron physics calculation model, calls and runs the neutron physics calculation program and obtains the core fission power shape distribution, fuel consumption nuclide information and core critical state information; the power conversion module receives the fission power distribution generated by the neutron physics simulation The data is arranged and converted into a data format that can be recognized and applied by the thermal-hydraulic program, and then passed to the thermal-hydraulic simulation module. After receiving the data, the thermal-hydraulic simulation module calculates and outputs the fuel temperature field, coolant temperature field and coolant density field. The fuel consumption nuclide update module converts the data according to the nuclide information generated by the neutron physics program and adds it back into the neutron physics calculation model. The control drum position update module obtains the core critical state information according to the neutron physics calculation program, solves the control drum value function, and updates the control drum position in the neutron physics calculation model according to the coordinates and diameter of the control drum.

2. The multi-physics field coupling system for a micro reactor according to claim 1, characterized in that: The microreactor comprises: a reflector layer and hexagonal fuel disposed therein, six control drum units for reactivity control located outside the fuel layer, a plurality of lead-bismuth coolants located inside the fuel layer, and a control rod for emergency shutdown, wherein the six control drum units are divided into two groups and are rotated by independent drive mechanisms. The control drum unit includes a main body material and a neutron absorbing material located outside it. The control drum unit is formed by Boolean operations on two circles and two lines. In the neutron physics model file, the control drum is represented as: -R_1∩+R_2∩-L_1∩-L_2, made of B4C! (-R_1∩+R_2∩-L_1∩-L_2), made of BeO. The negative sign (-) indicates inside the circle or below the line, the positive sign (+) indicates outside the circle or above the line, and ! indicates negation. Therefore, -R_1∩+R_2∩-L_1∩-L_2 represents the intersection of the inside of circle R_1, the outside of circle R_2, the bottom of line L_1, and the bottom of line L_2, resulting in the black portion in Figure 4. The black portion is made of B4C! (-R_1∩+R_2∩-L_1∩-L_2) is expressed as the part excluding the black part, and the material is BeO.

3. The multi-physics field coupling system for a micro reactor according to claim 2, characterized in that: When the center coordinates of control drum unit 1 are (0, D), the arc angle of the neutron absorber is θ, the angle of the control drum facing the core is β, the radius of circle R_1 is R, and the radius of circle R_2 is r, then: The control equations of the control drum unit No. 1 include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: 2 + (yD) 2 =R 2 ; The equation of circle R_2 is: 2 +(yD) 2 =r 2 The above four equations can be used to obtain the control drum facing the core at any angle. The control equations of the control drum unit No. 2 include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: The equation of circle R_2 is: The control equations of the control drum unit No. 3 include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: The equation of circle R_2 is: The control equations of the No. 4 control drum unit include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: 2 + (y+D) 2 =R 2 ; The equation of circle R_2 is: 2 +(y+D) 2 =r 2 ; The control equations of the No. 5 control drum unit include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: The equation of circle R_2 is: The control equations of the No. 6 control drum unit include: The equation of the straight line L_1 is: The equation of line L_2 is: The equation of circle R_1 is: The equation of circle R_2 is:

4. A multi-physics field coupling method based on the system according to any one of claims 1 to 3, characterized in that: include: Step 1) Based on the neutron physics program, the neutron macroscopic cross sections of the fuel and coolant are obtained according to the initial fuel temperature, fuel density, coolant temperature, and coolant density. Then, the neutron transport equation and the burnup equation are solved to calculate the neutron fission power, the effective breeding factor of the core, and the distribution and ratio of the burnup nuclides. Step 2) converting the neutron fission rate into power distribution, thereby obtaining the power of each grid; Step 3) Assign the power of each grid to the thermal grid, and obtain the fuel temperature, fuel density, coolant temperature, and coolant density by solving the mass conservation equation, momentum conservation equation, and energy conservation equation; Step 4) When the temperature field and density field of the fuel and coolant obtained in step 3) converge, proceed to step 5); otherwise, update the fuel temperature, coolant temperature, and coolant density in the neutron physics model file and return to step 1); Step 5) When the preset burnup is reached, the calculation is terminated, and the effective growth factor, power distribution, fuel temperature, coolant temperature, coolant density, and control drum position change information of the reactor over the entire burnup life are obtained; if the preset number of burnup steps has not been reached, step 6) is executed to continue the calculation; Step 6) Run the burnup update module and the control drum position update module, and update the fuel temperature, coolant temperature and coolant density in the neutron physics model file and perform the next burnup calculation.

5. The multi-physics field coupling method according to claim 4, wherein include: Step 1) Based on the neutron physics simulation module, according to the initial fuel temperature Fuel density Coolant temperature Coolant density By evaluating the nuclear database, the neutron macroscopic cross section of the fuel and coolant is obtained, a neutron physics model file is generated, and the neutron physics program is called to solve the neutron transport equation and the burnup equation to calculate the fission power. and burnup nuclide files Where: p is the p-th burnup, q is the number of iterations within the p-th burnup step, and g is the grid number; Step 2) Convert the fission power to Convert to power distribution Since the neutron fission power obtained by the neutron program statistics is only a shape distribution, it needs to be converted into a power distribution. The sum of the fission power and the sum of the power are linearly related, and the coefficient S is obtained, and then That is, the power of each grid; Step 3) Use the thermal hydraulic simulation module to calculate the power of each grid Assign the thermal grid, and through the thermal hydraulic simulation module, load the power of each grid calculated by the power conversion module into the model file of the thermal hydraulic program, call the thermal hydraulic program, solve the mass conservation equation, momentum conservation equation and energy conservation equation, and obtain the fuel temperature Fuel density Coolant temperature and coolant density Step 4) When the temperature field and density field of the fuel and coolant obtained in step 3) converge, proceed to step 5); otherwise, set the fuel temperature to Fuel density Coolant temperature Coolant density Import the neutron physics simulation module and return to step 1); The convergence mentioned here refers to: in: are the residuals of fuel temperature, coolant temperature, and coolant density, respectively. Calculate the maximum residuals of the fuel temperature, coolant temperature, and coolant density allowed respectively; Step 5) When the burnup time reaches a preset value, the calculation is terminated, and the effective growth factor, power distribution, fuel temperature, coolant temperature, coolant density, and control drum position change information of the reactor throughout the burnup life are obtained. Otherwise, step 6) is executed. Step 6) The burnup nuclide update module updates the neutron physics model file through the neutron physics simulation module based on the burnup nuclide file generated in step 1), and calls the neutron physics program to perform the next burnup calculation, specifically: according to the fuel temperature obtained in step 4) Fuel density Coolant temperature and coolant density And the nuclide data obtained in step 6), and call the control drum position update module to adjust the angle of the control drum, run the neutron physics simulation module, generate a new neutron physics model file, run the neutron physics program, perform the p+1 step burnup calculation, and proceed to step 1).

6. The multi-physics field coupling method according to claim 5, characterized in that: The effective proliferation factor refers to the ratio of the number of neutrons in the new generation of the reactor core to the number of neutrons in the immediately preceding generation. When it is greater than 1.0, the reactor is in a supercritical state and the power continues to rise. When it is 1.0, the power of the reactor remains unchanged and is in a stable state. When it is less than 1.0, the power of the reactor continues to decay.

7. The multi-physics field coupling method according to claim 5, characterized in that: The conversion into power distribution refers to: extracting the fission power of each grid in the Tally format file calculated by the neutron physics program, and then converting it into the power of each grid through step 2) to form a format file that can be read by the thermal hydraulic simulation module; The data conversion is to convert the nuclide information into nuclide data.

8. The multi-physics field coupling method according to claim 4 or 5, characterized in that: The neutron transport equation is: in: For the location The energy is E and the direction of motion is The neutron angular fluence rate, For the location The macroscopic cross section of neutrons with energy E is determined by the fuel temperature. Fuel density Coolant temperature Coolant density By evaluating the nuclear database, For the location The energy is E and the direction of motion is A scattered neutron source, For the location The energy is E and the direction of motion is When solving the neutron transport equation for an isotropic fission neutron source, given the neutron macroscopic cross section and the fission neutron source, the neutron flux rate is calculated. The neutron flux rate is multiplied by the neutron fission cross section. Get the neutron fission rate The fuel consumption equation is: Where: N i (t) is the concentration of nuclide i at time t, l ij is the proportion of nuclide i formed by radioactive decay in nuclide j, λ j is the decay constant of nuclide j, N j is the concentration of nuclide j, f ik is the proportion of nuclide i formed by neutron reaction in nuclide k, φ is the average neutron fluence rate, σ k is the microscopic cross section of nuclide k, N k is the concentration of nuclide k, λ i is the decay constant of nuclide i, σ i is the microscopic cross section of nuclide i, N i represents the concentration of nuclide i, and M in the burnup equation represents the number of nuclides of interest.

9. The multi-physics field coupling method according to claim 4 or 5, characterized in that: The mass conservation equation is: The momentum conservation equation is: The energy conservation equation is: in: are the velocities in the x, y, and z directions, μ is the viscosity, and f x , f y , f z is the external force on the grid, T is the grid temperature, λ is the thermal conductivity of the fluid, c p is the constant pressure specific heat capacity of the fluid, S T is the viscous dissipation term.

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