CN116956770B - Multi-physical field coupling method for reactor core of heat pipe reactor - Google Patents
Multi-physical field coupling method for reactor core of heat pipe reactor Download PDFInfo
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- 238000010168 coupling process Methods 0.000 title claims abstract description 38
- 238000004364 calculation method Methods 0.000 claims abstract description 92
- 230000008878 coupling Effects 0.000 claims abstract description 28
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- 238000002076 thermal analysis method Methods 0.000 claims abstract description 21
- 238000004458 analytical method Methods 0.000 claims abstract description 16
- 238000013461 design Methods 0.000 claims abstract description 6
- 239000000463 material Substances 0.000 claims abstract description 5
- 239000000446 fuel Substances 0.000 claims description 48
- 238000000034 method Methods 0.000 claims description 22
- 239000011159 matrix material Substances 0.000 claims description 19
- 230000035755 proliferation Effects 0.000 claims description 13
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- 238000010438 heat treatment Methods 0.000 description 1
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- G—PHYSICS
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- G06F30/20—Design optimisation, verification or simulation
- G06F30/28—Design optimisation, verification or simulation using fluid dynamics, e.g. using Navier-Stokes equations or computational fluid dynamics [CFD]
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- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F30/00—Computer-aided design [CAD]
- G06F30/20—Design optimisation, verification or simulation
- G06F30/23—Design optimisation, verification or simulation using finite element methods [FEM] or finite difference methods [FDM]
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- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F2119/00—Details relating to the type or aim of the analysis or the optimisation
- G06F2119/08—Thermal analysis or thermal optimisation
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- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F2119/00—Details relating to the type or aim of the analysis or the optimisation
- G06F2119/14—Force analysis or force optimisation, e.g. static or dynamic forces
Abstract
The invention discloses a multi-physical field coupling method for a reactor core of a heat pipe reactor, which comprises the following steps: firstly, a pretreatment part comprises establishing a reactor core geometric model, dividing grids aiming at the geometric model, and setting physical parameters and boundary conditions; secondly, program initialization setting, setting an initial temperature of core material, and carrying out core initial neutron physical calculation to obtain the power and distribution of the core, so as to provide input for loop iteration; then, carrying out thermal analysis calculation according to the initial power distribution of the reactor core, obtaining the temperature distribution of the reactor core through the thermal analysis calculation, and carrying out mechanical analysis calculation according to the temperature distribution result of the reactor core; and updating the deformation geometry obtained by thermal analysis and calculation and mechanical analysis and calculation into a neutron physical solving input card, correcting the reactor core geometry, material parameters and nuclear data, carrying out neutron physical calculation again, and entering the next iteration until the convergence condition is met. And finally obtaining neutron parameters, temperature distribution and stress strain distribution of the reactor core which are calculated by taking the nuclear thermal coupling into consideration. The invention lays a foundation for the design and optimization of the heat pipe reactor.
Description
Technical Field
The invention relates to the field of nuclear reactor core design and optimization analysis, in particular to a multi-physical field coupling method for a heat pipe reactor core.
Background
The heat pipe reactor has potential applicability and application value in the aspects of deep space exploration, space power supply, star surface energy supply, deep sea small-sized propeller, land-based motor-driven nuclear power supply, nuclear battery and the like due to the good inherent safety characteristics, modularized design concept and the like.
The reactor core of the solid heat pipe reactor with compact structure relates to the physical process of interaction of a plurality of physical fields such as reactor physics, thermal engineering, mechanical mechanics and the like in the operation process. The reactor core of the solid heat pipe reactor has high operating temperature, and the thermal expansion effect of the reactor core material is obvious at high temperature, so that larger deformation can be generated; the non-uniform heating and deformation enable the reactor core to have a dynamic geometric boundary, and the neutron and mechanical characteristics of the material can be changed, so that the neutron physical process of the reactor core is affected; the core power distribution in turn affects the core temperature distribution and thermal expansion. Through the core steady-state nuclear-thermal-force multi-physical field coupling numerical simulation, the interaction mechanism among neutron physics, core heat transfer and matrix deformation of the heat pipe reactor core is revealed, and the coupling phenomenon among all physical fields is researched, so that the method is an important method for researching the heat pipe reactor characteristics.
The method comprises the steps of modeling and calculating main multi-physical-field coupling processes such as neutron transport, heat and mass transfer, expansion and the like of a reactor core of a heat pipe reactor by using existing commercial software, grasping the influence rule and degree of key parameters in each physical field on the physical process, exploring the coupling phenomenon and the coupling mechanism between the physical fields, and researching the influence of different parameters and boundary conditions on each physical process. And writing a program external coupling interface, realizing data transmission among software, researching a discrete grid mapping method among different calculation modules, and forming a heat pipe pile multi-physical field external coupling analysis platform. Based on the coupling analysis platform, a mathematical physical model of main physical phenomena such as neutron transport, heat exchange of the heat pipe, heat conduction of the reactor body, mechanical stress and the like and the coupling process of the main physical phenomena is established, and a foundation is laid for the development of a multi-physical field coupling analysis program of the reactor core of the heat pipe reactor.
Disclosure of Invention
In order to overcome the problems in the prior art, the invention aims to provide a multi-physical field coupling method for a reactor core of a heat pipe reactor, which reduces the coupling modeling difficulty and provides a basis for realizing the physical and thermal coupling analysis of the reactor core of the heat pipe reactor.
In order to achieve the above purpose, the invention adopts the following technical scheme:
the method is an external coupling method based on a Monte Carlo neutron transport program and an ANSYS Workbench multi-physical field coupling analysis program, and comprises the following steps:
step 1: the method comprises the following steps of calculating pretreatment settings, performing initial settings required by core neutron physical calculation, core thermal analysis calculation and core mechanics analysis calculation, and specifically comprising the following steps:
step 1-1: executing a nuclear data processing program, acquiring Doppler broadening sections and thermal scattering correction of each isotope at intervals of 50K within a temperature range of 300K to 2000K, and constructing a section library required by the vertical neutron physical calculation;
step 1-2: for a given heat pipe reactor core structure, establishing a geometric model of the reactor core by adopting three-dimensional modeling software, generating neutron physical calculation input files, and carrying out hexahedral mesh division on the geometric model of the reactor core by adopting mesh division software to obtain a mesh model;
step 1-3: importing the grid model in the step 1-2 into computational fluid dynamics software, carrying out initialization solution setting on the computational fluid dynamics software, setting the initial temperature of materials, and setting the initial maximum temperature of a computational domain as T 0 Setting the initial effective multiplication coefficient of the reactor core as k 0 Setting boundary conditions, and compiling a user-defined function UDF; setting a monitor to monitor the highest temperature of a calculation domain, and using the highest temperature as a judgment basis for subsequent calculation convergence;
step 2: performing neutron physical calculation in the reactor core to generate a calculation output file, wherein the output file contains calculation results of the effective proliferation coefficient, the power and the distribution of the reactor core;
step 3: the user-defined function control is used for carrying out core thermal analysis calculation to obtain the temperature distribution of the core fuel, the matrix and the heat pipe, and the specific steps are as follows:
step 3-1: analyzing an output file of neutron physical calculation in a reactor core by using a user-defined function of computational fluid dynamics software, traversing the output file, reading the count of fuel rod cells, and converting to obtain the power of the fuel rod cells, wherein the conversion formula is as follows:
wherein: p (P) Cell unit For the power of a certain cell of a fuel rod in the neutron physical calculation of a reactor core, P 0 For the design value of the reactor power, N is the total count, N i,j A count value for the cell;
step 3-2: using the power of the fuel rod cells obtained in the step 3-1 as the input of the core thermal analysis calculation, enabling the power of the fuel rod cells to be corresponding to the fuel rod grid cells by a source term function in a user-defined function, and then starting the core thermal analysis calculation;
step 3-3: when the monitor set in the step 1-3 monitors that the change of the highest temperature in the calculation domain is less than 1K, the calculation is considered to be converged, the core thermal analysis calculation is completed, and the average temperature of the fuel rod grid unit corresponding to the fuel rod grid unit and the whole matrix is output to the output file of the step 2, wherein the average temperature calculation formula is as follows:
wherein: t (r, z) is the grid temperature with the radial coordinate r and the axial coordinate z; v (V) i Is the volume of the grid cell; t (T) i Temperature values within the grid cells;
step 4: mapping the temperature distribution of the fuel rods and the matrix obtained by core thermal analysis calculation into a grid for core mechanical analysis calculation by an ANSYS Workbench multi-physical field coupling analysis program, starting the core mechanical analysis calculation to obtain stress strain and deformation of the fuel rods and the matrix, and outputting the average deformation of each fuel rod into an output file of the step 2;
step 5: the average temperature of the fuel rod grid units and the average temperature of the whole matrix output in the step 3-3 are utilized to update the temperatures of the fuel rods and the corresponding grid units of the matrix in the Monte Carlo neutron transport program input file, and the average deformation of each fuel rod output in the step 4 is utilized to update the geometric diameters of the corresponding fuel grid units in the Monte Carlo neutron transport program input file;
step 6: when the monitor set in the step 1-3 monitors the calculated domain maximum temperature and the set calculated domain initial maximum temperature T 0 When the difference is smaller than 1K, the effective proliferation coefficient of the reactor core output in the step 2 is equal to the set initial effective proliferation coefficient K of the reactor core 0 The difference is less than 10 -5 The coupling calculation is considered to be converged, and the calculation is completed; if neither condition is satisfied, the set initial maximum temperature T of the calculation domain is set 0 The maximum temperature of the monitoring calculation domain is equal to the monitor, and the set initial effective proliferation coefficient k of the reactor core is set 0 And (3) the effective multiplication coefficient of the reactor core output in the step (2) is equal, and the steps (2) to (6) are repeated.
The beneficial effects are that:
compared with common physical and thermal coupling treatment, the invention has the following beneficial effects: 1) The method provided by the invention calls each sub-calculation module to carry out the calculation of the core-heat-force coupling of the heat pipe reactor, has definite division of work and high calculation precision, can complete the whole coupling analysis process, reduces the use difficulty of users, and is convenient for the users.
2) The invention fully considers the coupling effect of neutron physics, thermal engineering and mechanical mechanics in the core operation process of the heat pipe pile, provides a definite nuclear-thermal-force coupling strategy and a data mapping method among power, temperature and deformation, and provides thought for the coupling analysis of multiple physical fields of the solid heat pipe pile.
3) In the coupling process, the method adopts the convergence criterion of calculating the double convergence of the highest temperature of the domain and the effective proliferation coefficient of the reactor core, so that the coupling calculation accuracy can be ensured.
Drawings
FIG. 1 is a flow chart of a method for multi-physical field coupling of a heat pipe reactor core according to the present invention.
Detailed Description
The method of the present invention for multi-physical field coupling of a hot pipe reactor core is described in further detail below with reference to the flow chart of FIG. 1:
step 1: the method comprises the following steps of calculating pretreatment settings, performing initial settings required by core neutron physical calculation, core thermal analysis calculation and core mechanics analysis calculation, and specifically comprising the following steps:
step 1-1: executing a nuclear data processing program NJOY, acquiring Doppler broadening sections and thermal scattering correction of each isotope at intervals of 50K within a temperature range of 300K to 2000K, and constructing a section library required by neutron physical calculation;
step 1-2: for a given heat pipe reactor core structure, establishing a geometric model of the reactor core by adopting three-dimensional modeling software Solidworks, generating neutron physical calculation input files, and carrying out hexahedral mesh division on the geometric model of the reactor core by adopting mesh division software CFD-ICEM to obtain a mesh model;
step 1-3: importing the grid model in the step 1-2 into Fluent, initializing, solving and setting Fluent software according to a software use instruction, setting initial temperature of materials, and setting initial highest temperature of a computing domain as T 0 Setting up a stackThe initial effective multiplication coefficient of the core is k 0 Setting boundary conditions, and compiling a user-defined function UDF; when the monitor is set, the monitor is set to monitor the highest temperature of the calculation domain and is used as a judgment basis for the convergence of the subsequent calculation;
step 2: performing neutron physical calculation in the reactor core to generate a calculation output file, wherein the output file contains calculation results of the effective proliferation coefficient, the power and the distribution of the reactor core;
step 3: the user-defined function control carries out core thermal analysis calculation to obtain the temperature distribution of the core fuel, the matrix and the heat pipe, and the method comprises the following steps:
step 3-1: analyzing an output file of neutron physical calculation in a reactor core by using a user-defined function of Fluent, traversing the output file, reading the count of fuel rod cells, and converting to obtain the power of the fuel rod cells, wherein the conversion formula is as follows:
wherein: p (P) Cell unit For the power of a certain cell of a fuel rod in the neutron physical calculation of a reactor core, P 0 For the design value of the reactor power, N is the total count, N i,j A count value for the cell;
step 3-2: using the power of the fuel rod cells obtained in the step 3-1 as the input of the core thermal analysis calculation, enabling the power of the fuel rod cells to be corresponding to the fuel rod grid cells by a source term function in a user-defined function, and then starting the core thermal analysis calculation;
step 3-3: when the monitor set in the step 1-3 monitors that the change of the highest temperature in the calculation domain is less than 1K, the calculation is considered to be converged, the core thermal analysis calculation is completed, and the average temperature of the fuel rod grid unit corresponding to the fuel rod grid unit and the whole matrix is output to the output file of the step 2, wherein the average temperature calculation formula is as follows:
wherein: t (r, z) is the grid temperature with the radial coordinate r and the axial coordinate z; v (V) i Is the volume of the grid cell; t (T) i Temperature values within the grid cells;
step 4: mapping the temperature distribution of the fuel rods and the matrix obtained by core thermal analysis calculation into a grid for core mechanical analysis calculation by an ANSYS Workbench multi-physical field coupling analysis program, starting the core mechanical analysis calculation to obtain stress strain and deformation of the fuel rods and the matrix, and outputting the average deformation of each fuel rod into an output file of the step 2;
step 5: and (3) respectively updating the temperatures of the fuel rods and corresponding cells of the matrix in the Monte Carlo neutron transport program input file by using the average temperature of the fuel rod grid cells and the average temperature of the whole matrix output in the step (3-3), and updating the geometric diameters of the corresponding fuel cells in the Monte Carlo neutron transport program input file by using the average deformation of each fuel rod output in the step (4).
Step 6: when the monitor set in the step 1-3 monitors the calculated domain maximum temperature and the set calculated domain initial maximum temperature T 0 When the difference is smaller than 1K, the effective proliferation coefficient of the reactor core output in the step 2 is equal to the set initial effective proliferation coefficient K of the reactor core 0 The difference is less than 10 -5 The coupling calculation is considered to be converged, and the calculation is completed; if neither condition is satisfied, the set initial maximum temperature T of the calculation domain is set 0 The maximum temperature of the monitoring calculation domain is equal to the monitor, and the set initial effective proliferation coefficient k of the reactor core is set 0 And (3) the effective multiplication coefficient of the reactor core output in the step (2) is equal, and the steps (2) to (6) are repeated.
The invention is not described in detail in the field of technical personnel common knowledge. In the above steps, several pieces of software are involved, and the user can use the software or replace the software with similar software.
Claims (1)
1. The method is an external coupling method based on a Monte Carlo neutron transport program and an ANSYS Workbench multi-physical field coupling analysis program, and comprises the following steps:
step 1: the method comprises the following steps of calculating pretreatment settings, performing initial settings required by core neutron physical calculation, core thermal analysis calculation and core mechanics analysis calculation, and specifically comprising the following steps:
step 1-1: executing a nuclear data processing program, acquiring Doppler broadening sections and thermal scattering correction of each isotope at intervals of 50K within a temperature range of 300K to 2000K, and constructing a section library required by neutron physical calculation;
step 1-2: for a given heat pipe reactor core structure, establishing a geometric model of the reactor core by adopting three-dimensional modeling software, generating neutron physical calculation input files, and carrying out hexahedral mesh division on the geometric model of the reactor core by adopting mesh division software to obtain a mesh model;
step 1-3: importing the grid model in the step 1-2 into computational fluid dynamics software, carrying out initialization solution setting on the computational fluid dynamics software, setting the initial temperature of materials, and setting the initial maximum temperature of a computational domain as T 0 Setting the initial effective multiplication coefficient of the reactor core as k 0 Setting boundary conditions, and compiling a user-defined function UDF; setting a monitor to monitor the highest temperature of a calculation domain, and using the highest temperature as a judgment basis for subsequent calculation convergence;
step 2: performing neutron physical calculation in the reactor core to generate a calculation output file, wherein the output file contains calculation results of the effective proliferation coefficient, the power and the distribution of the reactor core;
step 3: the user-defined function control is used for carrying out core thermal analysis calculation to obtain the temperature distribution of the core fuel, the matrix and the heat pipe, and the specific steps are as follows:
step 3-1: analyzing an output file of neutron physical calculation in a reactor core by using a user-defined function of computational fluid dynamics software, traversing the output file, reading the count of fuel rod cells, and converting to obtain the power of the fuel rod cells, wherein the conversion formula is as follows:
wherein:P cell unit For the power of a certain cell of a fuel rod in the neutron physical calculation of a reactor core, P 0 For the design value of the reactor power, N is the total count, N i,j A count value for the cell;
step 3-2: using the power of the fuel rod cells obtained in the step 3-1 as the input of the core thermal analysis calculation, enabling the power of the fuel rod cells to be corresponding to the fuel rod grid cells by a source term function in a user-defined function, and then starting the core thermal analysis calculation;
step 3-3: when the monitor set in the step 1-3 monitors that the change of the highest temperature in the calculation domain is less than 1K, the calculation is considered to be converged, the core thermal analysis calculation is completed, and the average temperature of the fuel rod grid unit corresponding to the fuel rod grid unit and the whole matrix is output to the output file of the step 2, wherein the average temperature calculation formula is as follows:
wherein: t (r, z) is the grid temperature with the radial coordinate r and the axial coordinate z; v (V) i Is the volume of the grid cell; t (T) i Temperature values within the grid cells;
step 4: mapping the temperature distribution of the fuel rods and the matrix obtained by core thermal analysis calculation into a grid for core mechanical analysis calculation by an ANSYS Workbench multi-physical field coupling analysis program, starting the core mechanical analysis calculation to obtain stress strain and deformation of the fuel rods and the matrix, and outputting the average deformation of each fuel rod into an output file of the step 2;
step 5: the average temperature of the fuel rod grid units and the average temperature of the whole matrix output in the step 3-3 are utilized to update the temperatures of the fuel rods and the corresponding grid units of the matrix in the Monte Carlo neutron transport program input file, and the average deformation of each fuel rod output in the step 4 is utilized to update the geometric diameters of the corresponding fuel grid units in the Monte Carlo neutron transport program input file;
step 6: when the monitor set in step 1-3 monitors the highest temperature of the calculation domain and the settingCalculating the domain initial maximum temperature T 0 When the difference is smaller than 1K, the effective proliferation coefficient of the reactor core output in the step 2 is equal to the set initial effective proliferation coefficient K of the reactor core 0 The difference is less than 10 -5 The coupling calculation is considered to be converged, the calculation is completed, and neutron parameters, temperature distribution and stress strain distribution of the reactor core, which take account of the nuclear thermal coupling calculation, are obtained; if neither condition is satisfied, the set initial maximum temperature T of the calculation domain is set 0 The maximum temperature of the monitoring calculation domain is equal to the monitor, and the set initial effective proliferation coefficient k of the reactor core is set 0 And (3) the effective multiplication coefficient of the reactor core output in the step (2) is equal, and the steps (2) to (6) are repeated.
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