CN112906272B - Reactor steady-state physical thermal full-coupling fine numerical simulation method and system - Google Patents
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Abstract
The invention discloses a reactor steady-state physical thermal full-coupling fine numerical simulation method and a system, wherein the method comprises the following steps: step S1, establishing a reactor steady-state physical thermal coupling nonlinear equation set; and step S2, solving the nonlinear equation set by adopting a JFNK method to obtain the neutron flux, the power, the coolant temperature and the fuel rod temperature of the whole reactor core fuel grid cell level. The invention adopts a fine net difference method to disperse a neutron physical diffusion equation, adopts a single channel model to simulate a coolant flow heat transfer equation and adopts a finite volume method to disperse a cylinder heat conduction equation, establishes a nonlinear equation set for simulating the physical and thermal professions of a reactor, adopts a JFNK method to solve the nonlinear equation set, obtains three-dimensional physical and thermal parameters of a fuel grid element level of a whole reactor core, and improves the accuracy of the physical and thermal coupling calculation of the reactor.
Description
Technical Field
The invention belongs to the technical field of nuclear reactor core calculation, and particularly relates to a reactor steady-state physical and thermal full-coupling fine numerical simulation method and system, and a computer readable storage medium and computer equipment for storing and executing the method.
Background
Reactor physical-thermal coupling means that there is a significant feedback effect between physical quantities in the reactor that are physically and thermally specialized. The method specifically comprises the following steps: the fission power obtained by the physical professional calculation influences the fuel temperature, the moderator density and related physical parameters calculated by the thermal professional calculation, and the fuel temperature and the moderator density calculated by the thermal professional influence the section parameters of the physical professional, so that the fission power calculated by the physical professional is influenced, and the coupling process is shown in the attached figure 1.
In order to obtain more accurate reactor parameters in the design and safety analysis of the reactor, the coupling effect between physical and thermal engineering must be considered.
The physical and thermal coupling calculation method widely adopted at present is an operator splitting method, and the implementation process is shown in the attached figure 2: the physical equation and the thermal equation are independent from each other and are solved according to a certain sequence, and two professional data interactions are realized in a data interface mode. According to the iteration times and convergence conditions between the physics and the thermal engineering of a time step, the method can be divided into a simple operator splitting method, a semi-implicit operator splitting method and a Picard iteration method. The operator splitting method is essentially loose coupling, a physical equation and a thermal equation are decoupled and solved, only the first-order convergence rate is achieved, and the calculation efficiency and the convergence of the operator splitting method are difficult to guarantee for complex physical and thermal coupling problems.
In order to overcome the defects of the operator splitting method, a fully-coupled calculation for simultaneous solution of a physical equation and a thermal equation is proposed in recent years. Researchers have proposed a joint solution method for the physical thermal equation of neutron physics by using a block method and a single channel for thermal hydraulic power. However, because the grid of the block method is about a component level and has a large size, the operation of homogenization treatment exists when the grid is coupled with a thermal specialty, and only the calculation of the physical and thermal coupling effect of the component level can be realized, but the physical and thermal coupling effect of the fuel grid cell level in the component cannot be simulated more accurately.
Disclosure of Invention
The invention provides a reactor steady-state physical thermal full-coupling fine numerical simulation method, aiming at solving the problem of poor precision and reliability of the existing nuclear reactor steady-state physical thermal full-coupling numerical simulation technology. The invention adopts a fine net difference method to disperse neutron physical diffusion equation, a single channel model to simulate coolant flow heat transfer equation and a finite volume method to disperse cylinder heat conduction equation, establishes a nonlinear equation set for simulating reactor physics and thermal engineering specialties, and adopts a Jacbian-free Newton Krylov (JFNK) method to solve the nonlinear equation set, thereby obtaining three-dimensional steady state physical and thermal engineering parameters of the whole reactor core fuel grid level and improving the precision of reactor physics and thermal engineering coupling calculation.
The invention is realized by the following technical scheme:
the invention discloses a reactor steady-state physical thermal full-coupling fine numerical simulation method, which comprises the following steps:
step S1, establishing a reactor steady-state physical thermal coupling nonlinear equation set:
wherein f isφ(x)、fQ(x)、
And fλ(x) Neutron diffusion equation, neutron flux-Residual forms of additional equations of a power equation, a coolant energy steady state equation, a fuel heat conduction steady state equation, and a closed neutron diffusion equation; x is a solution vector;
and step S2, solving the nonlinear equation set by adopting a JFNK method to obtain the steady-state neutron flux, power, coolant temperature and fuel rod temperature of the whole reactor core fuel grid cell level.
Preferably, the steady-state neutron diffusion equation of the present invention adopts a central point difference format to perform spatial dispersion, and obtains a discrete expression of the g-th group neutron diffusion equation as follows:
wherein:
ap，g＝∑R，g，i，j，kΔVi，j，k-aw，g-ae，g-an，g-as，g-ab，g-at，g
wherein phi is neutron flux; g is the total number of energy groups; d is a diffusion coefficient; sigmag′→gIs a diffusion cross section from the g 'th group to the g' th group; sigmaRTo remove cross-sections; χ is a neutron fission spectrum; upsilon sigmafCreating a cross-section for neutron fission; k is a radical ofeffEffective multiplication factor; a isw、ae、an、as、at、abAnd apRespectively are coefficients of all directions when the central mesh points are different; Δ x is the x-direction grid spacing; Δ y is the y-direction grid spacing; Δ z is the grid spacing in the z direction; Δ V is the grid volume; g is the g energy group; g 'is the g' th energy group; i. j and k are grid coordinates in x, y and z directions, respectively. The neutron diffusion equation provided by the invention adopts a differential format to carry out spatial dispersion, and can promote the steady-state physical thermal full-coupling calculation from an assembly level to a fuel rod level.
Preferably, the steady-state neutron flux-power equation of the present invention adopts a central point difference format to perform spatial dispersion, and the discrete expression of the neutron flux-power equation is obtained as follows:
in the formula, Qi,j,kIs the actual power of the grid (i, j, k);
flux-power normalization coefficients; kappa sigmafIs an energy fission cross section; g is the total number of energy groups; Δ V is the grid volume; i. j and k are grid coordinates in the x direction, the y direction and the z direction respectively; phi is neutron flux; g is the g-th energy group.
Preferably, the coolant energy steady-state equation of the present invention adopts a single-channel model, and spatially disperses by a finite volume method, and the discrete expression of the coolant energy steady-state equation is obtained by:
in the formula, TcIs the coolant temperature; w is the coolant mass flow rate; cpThe specific heat capacity is constant pressure of the coolant; qi,j,kIs the actual power of the grid (i, j, k); i. j and k are grid coordinates in x, y and z directions, respectively.
Preferably, the discrete expression of the fuel heat conduction steady state equation of the present invention is:
in the formula, TfIs the fuel temperature; q is the volumetric heat release rate; r is the fuel rod radius; h is the convective heat transfer coefficient of the coolant; lambda is the heat conductivity coefficient of the fuel rod; t iscIs the coolant temperature; i. j and k are grid coordinates in x, y and z directions, respectively.
Preferably, the expression of the additional equation of the closed neutron diffusion equation of the present invention is:
wherein G is the total number of energy groups; phi is the neutron flux.
Preferably, step S2 of the present invention specifically includes:
converting the system of nonlinear equations into a linear expression:
J(xk)δxk＝-F(xk)
wherein J is the attic ratio matrix of f (x) 0; δ x is the argument increment; x is the number ofkIs an independent variable;
solving the linear expression of the nonlinear equation set by adopting a JFNK method, wherein the specific solving process is as follows:
step S21, constructing an Attic ratio vector or an approximate vector of the Attic ratio vector; the Acigure ratio vector is the product of an Acigure ratio matrix and a Krylov subspace basis vector;
step S22, solving a nonlinear equation set J (x) by adopting a generalized minimum residual error method GMRESk)δxk＝-F(xk) Obtaining the independent variable increment delta xk；
Step S23, updating the argument xk+1＝xk+δxkCalculating environmental parameters calculated by the independent variables;
step S24, repeating the steps S21-S23 until | | | F (x)k+1)||≤eps，||F(xk+1) And | | is a second-order norm of the residual error of the nonlinear equation set, and eps is convergence precision.
On the other hand, the invention also provides a reactor steady-state physical thermal full-coupling fine numerical simulation system, which comprises a model construction module, an analysis module and an output module;
the model building module is used for building a reactor steady-state physical thermal coupling nonlinear equation system:
wherein f isφ(x)、fQ(x)、
And fλ(x) Residual forms of additional equations, which are a steady-state neutron diffusion equation, a neutron flux-power equation, a coolant energy steady-state equation, a fuel heat conduction steady-state equation, and a closed neutron diffusion equation, respectively; x is a solution vector;
the analysis module solves the nonlinear equation set by adopting a JFNK method to obtain the steady-state neutron flux, power, coolant temperature and fuel rod temperature of the fuel grid cell level of the whole reactor core;
and the output module is used for outputting the three-dimensional steady-state physical parameters and the thermal parameters of the whole reactor core fuel grid cell level obtained by the solution of the analysis module.
The invention also proposes a computer device comprising a memory and a processor, the memory storing a computer program, the processor implementing the steps of the above-mentioned method of the invention when executing the computer program.
The invention also proposes a computer-readable storage medium, on which a computer program is stored which, when being executed by a processor, carries out the steps of the above-mentioned method of the invention.
The invention has the following advantages and beneficial effects:
according to the objective coupling rule among reactor physical thermal engineering, a reactor steady-state physical thermal engineering coupling nonlinear equation set is established, and a JFNK method is adopted to solve the nonlinear equation set to obtain neutron flux, power, coolant temperature and fuel rod temperature; the JFNK method adopted by the invention keeps the second-order convergence rate of the Newton iteration method, ensures the calculation efficiency and precision of the steady-state physical thermal fully-coupled calculation, can overcome the defects of low convergence rate and thick grid of a node method in the traditional operator splitting method, effectively improves the calculation efficiency and fineness of the reactor steady-state physical thermal fully-coupled calculation, and improves the accuracy and reliability of the reactor steady-state physical thermal fully-coupled numerical simulation output data.
The method can be used for fine steady-state physical thermal coupling calculation of typical pressurized water reactors such as Hualong I and the like.
Drawings
The accompanying drawings, which are included to provide a further understanding of the embodiments of the invention and are incorporated in and constitute a part of this application, illustrate embodiment(s) of the invention and together with the description serve to explain the principles of the invention. In the drawings:
FIG. 1 is a schematic diagram of a nuclear reactor physical-thermal coupling process.
Fig. 2 is a flow chart of a conventional physical-thermal coupling calculation method (operator splitting method).
FIG. 3 is a schematic flow chart of the method of the present invention.
FIG. 4 is a schematic flow chart of iterative solution of a steady-state physical-thermal fully-coupled equation set by adopting a JFNK method.
FIG. 5 is a schematic diagram of a computer device according to the present invention.
Fig. 6 is a schematic block diagram of the system of the present invention.
Detailed Description
In order to make the objects, technical solutions and advantages of the present invention more apparent, the present invention is further described in detail below with reference to examples and accompanying drawings, and the exemplary embodiments and descriptions thereof are only used for explaining the present invention and are not meant to limit the present invention.
Example 1
Compared with the defects of the traditional physical thermal coupling calculation method, the method provided by the embodiment provides a reactor steady-state physical thermal fully-coupled fine numerical simulation method, the method provided by the embodiment adopts a fine net difference method to disperse a neutron physical diffusion equation, adopts a single-channel model to simulate a coolant flow heat transfer equation, adopts a finite volume method to disperse a cylindrical heat conduction equation, establishes a nonlinear equation set for simulating the physical and thermal specialties of the reactor, adopts a JFNK method to solve the nonlinear equation set, obtains three-dimensional physical parameters and thermal parameters of a fuel grid element level of the whole reactor core, and improves the accuracy of the reactor physical thermal coupling calculation.
As shown in fig. 3, the method of this embodiment specifically includes:
step S1, establishing a reactor steady-state physical thermal coupling nonlinear equation set, wherein the expression is as follows:
in the formula (f)φ(x)、fQ(x)、
And fλ(x) Respectively, are residual error forms of steady-state neutron diffusion equation, flux-power equation, coolant energy equation, fuel heat conduction equation and additional closed equation. F (x) is a simultaneous steady-state physical-thermal coupling calculation equation set, and x is a solution vector and is neutron flux and work respectivelyRate, coolant temperature and fuel temperature, characteristic values.
In this embodiment, a central point difference format is adopted for the steady-state multi-group neutron diffusion equation to perform spatial dispersion, and the dispersion expression of the g-th group neutron diffusion equation is as follows:
wherein,
ap,g＝∑R,g,i,j,kΔVi,j,k-aw,g-ae,g-an,g-as,g-ab,g-at,g (9)
in the formula,
phi-neutron flux; g is the total number of energy groups; d-diffusion coefficient; sigmag'→g-diffusion cross-section of group g' to group g; sigmaR-removing the cross section; chi-neutron fission spectrum; upsilon sigmaf-the neutron fission produces a cross section; k is a radical ofeff-an effective multiplication factor; a isw、ae、an、as、at、abAnd ap-coefficients of each direction when the central dots are differentiated; Δ x — grid spacing in the x direction; Δ y-y-direction grid spacing; Δ z-direction grid spacing; Δ V-mesh volume; g-the g th energy group; g '-the g' th energetics; i. j, k-grid coordinates in x, y, z directions, respectively.
Writing equation (2) to the matrix expression form:
MΦ＝λFΦ (10)
where λ is a characteristic value, λ is 1/keff(ii) a Phi is neutron flux in a vector form, which uniformly describes G groups of neutrons distributed according to a space grid, the energy groups are arranged according to 1-G, the space is sequentially arranged from left to right according to the lower left corner of the region, and the neutron flux in the first row starts from the second row after the first row is finished until the whole space. The expression form is as follows:
the expression form of the M matrix is:
in the formula MgThe g-th group neutron diffusion equation is expressed by adopting a matrix formed by all direction coefficients when the central lattice point difference is adopted, and M is used for solving one-dimensional, two-dimensional and three-dimensional problemsgThree diagonal, five diagonal and 7 diagonal matrices, respectively. The form can be simply exemplified as follows:
the expression form of the F matrix is:
in this embodiment, the flux-power return equation is spatially discretized in a central point difference format, and the discretization expression is as follows:
in the formula, Qr-actual reactor power;
-flux-power normalization coefficients; kappa sigmaf-an energy fission cross section; I. j, K-total number of grids in the x, y, and z directions, respectively.
Calculating the actual power of each grid by using the flux-power normalization coefficient, and performing spatial dispersion by using a central point difference format, wherein the expression is as follows:
in the formula,
Qi,j,k-representing the actual power of the grid (i, j, k).
Writing equation (16) to the form of a matrix:
in the formula, Eg-identifying a matrix for the power produced by the group g neutrons, the matrix being a two-dimensional single diagonal matrix whose expression is:
in this embodiment, a single-channel model is adopted for the steady-state equation of the coolant energy, and the finite volume method is spatially adopted for dispersion, where the discrete expression is:
in the formula,
Tc-the coolant temperature; w-coolant mass flow rate; cp is the constant pressure specific heat capacity of the coolant.
Equation (19) is written in matrix form as:
STc＝RQ (20)
wherein S is a coolant temperature coefficient matrix which is a two-diagonal matrix; r is a coefficient matrix of temperature increment, which is a diagonal matrix;
in this embodiment, a discrete expression of the fuel heat conduction steady-state equation is established, and the discrete expression does not consider the fuel rod axial heat conduction and is:
in the formula, Tf-the temperature of the fuel; q is the volume heat release rate,
r is fuel rod radius; h-convective heat transfer coefficient of coolant; lambda is the fuel rod thermal conductivity.
Equation (21) is written in matrix form as:
Tf＝Tc+Gq (22)
in the formula,
g is a coefficient matrix of the volume heat release rate, and is a single diagonal matrix.
This embodiment establishes an additional equation of the closed neutron diffusion equation, expressed in the form of:
the expression form of the multi-group is as follows:
and step S2, solving the nonlinear equation set by adopting a JFNK method to obtain the neutron flux, the power, the coolant temperature and the fuel rod temperature of the whole reactor core fuel grid cell level.
This example uses Newton's method to apply equation (1) at the current iteration point xkThe Taylor expansion is:
F(xk+1)＝F(xk)+F'(xk)(xk+1-xk)+higher-oerderterms (25)
ignoring higher order terms, assuming xk+1Is the exact solution of f (x) 0, equation (25) becomes:
J(xk)δxk＝-F(xk) (26)
in the formula,
j-the accharance vector of f (x) 0; δ x — argument increment.
In this embodiment, the JFNK method is adopted to solve the above equation (26), and the solving process is specifically shown in fig. 4, and includes:
in step S21, an attic ratio vector is constructed.
When the JFNK method is adopted to solve the nonlinear equation set, the Acigure ratio matrix does not need to be constructed explicitly, but the Acigure ratio vector needs to be formed. The present embodiment provides two methods to construct the Achates ratio vector.
The method comprises the following steps: construction of Accord vector (display construction Accord matrix) by analytical method
Constructing an Accord matrix for formula (24), the expression is:
in the formula, I is a unit diagonal matrix.
The Acigure vector is the product of the Acigure matrix and the Krylov subspace basis vector, and is expressed as:
the method 2 comprises the following steps: constructing an approximation vector of the Accord vector
An approximate vector of the Acertian vector is constructed by using finite difference, and the form is as follows:
where ε is a perturbation parameter, typically the square root of the precision of the floating point number of the machine.
Step S22: solving a linear equation set J (x) by adopting a GMRES method of a generalized minimum residual error methodk)δxk＝-F(xk) Obtaining the independent variable increment delta xk；
Step S23: updating the argument xk+1＝xk+δxkCalculating environmental parameters calculated by the independent variables;
step S24: repeating the steps S21-S23 until | | | F (x)k+1)||≤eps，||F(xk+1) And | | is a second-order norm of the residual error of the nonlinear equation set, and eps is convergence precision.
The embodiment also provides a computer device for executing the method of the embodiment.
As shown particularly in fig. 5, the computer device includes a processor, a memory, and a system bus; various device components including a memory and a processor are connected to the system bus. A processor is hardware used to execute computer program instructions through basic arithmetic and logical operations in a computer system. Memory is a physical device used for temporarily or permanently storing computing programs or data (e.g., program state information). The system bus may be any of several types of bus structures including a memory bus or memory controller, a peripheral bus, and a local bus. The processor and the memory may be in data communication via a system bus. Including read-only memory (ROM) or flash memory (not shown), and Random Access Memory (RAM), which typically refers to main memory loaded with an operating system and computer programs.
Computer devices typically include a storage device. The storage device may be selected from a variety of computer readable media, which refers to any available media that can be accessed by a computer device, including both removable and non-removable media. For example, computer-readable media includes, but is not limited to, flash memory (micro SD cards), CD-ROM, Digital Versatile Disks (DVD) or other optical disk storage, magnetic cassettes, magnetic tape, magnetic disk storage or other magnetic storage devices, or any other medium which can be used to store the desired information and which can be accessed by a computer device.
A computer device may be logically connected in a network environment to one or more network terminals. The network terminal may be a personal computer, a server, a router, a smart phone, a tablet, or other common network node. The computer apparatus is connected to the network terminal through a network interface (local area network LAN interface). A Local Area Network (LAN) refers to a computer network formed by interconnecting within a limited area, such as a home, a school, a computer lab, or an office building using a network medium. WiFi and twisted pair wiring ethernet are the two most commonly used technologies to build local area networks.
It should be noted that other computer systems including more or less subsystems than computer devices can also be suitable for use with the invention.
As described in detail above, the computer apparatus suitable for the present embodiment can perform the specified operations of a reactor steady-state physical thermal fully-coupled fine numerical simulation method. The computer device performs these operations in the form of software instructions executed by a processor in a computer-readable medium. These software instructions may be read into memory from a storage device or from another device via a local area network interface. The software instructions stored in the memory cause the processor to perform the method of processing group membership information described above. Furthermore, the present invention can be implemented by hardware circuits or by a combination of hardware circuits and software instructions. Thus, implementation of the present embodiments is not limited to any specific combination of hardware circuitry and software.
Example 2
Based on the above embodiment 1, the embodiment further provides a reactor steady-state physical-thermal fully-coupled fine numerical simulation system, which includes a model building module, an analysis module, and an output module.
The model construction module of the embodiment is used for establishing a reactor steady-state physical thermal coupling nonlinear equation set:
wherein f isφ(x)、fQ(x)、
And fλ(x) Residual forms of additional equations, respectively, of a steady-state neutron diffusion equation, a neutron flux-power equation, a coolant energy steady-state equation, a fuel heat conduction steady-state equation, and a closed neutron diffusion equation; x is a solution vector; the specific process is the same as that in embodiment 1, and is not described herein again.
The analysis module of the embodiment solves the nonlinear equation set by adopting a JFNK method to obtain the neutron flux, power, coolant temperature and fuel rod temperature of the fuel cell level of the whole reactor core; the specific process is the same as that in embodiment 1, and is not described herein again.
The output module of the embodiment is used for outputting the three-dimensional physical parameters and the thermal parameters of the fuel cell level of the whole reactor core obtained by the solution of the analysis module, so as to more accurately simulate the physical and thermal coupling effect of the fuel cell level in the assembly and provide more reliable and effective technical support for the design of the reactor core of the nuclear reactor.
The above-mentioned embodiments are intended to illustrate the objects, technical solutions and advantages of the present invention in further detail, and it should be understood that the above-mentioned embodiments are merely exemplary embodiments of the present invention, and are not intended to limit the scope of the present invention, and any modifications, equivalent substitutions, improvements and the like made within the spirit and principle of the present invention should be included in the scope of the present invention.
Claims (10)
1. A reactor steady-state physical thermal full-coupling fine numerical simulation method is characterized by comprising the following steps:
step S1, establishing a reactor steady-state physical thermal coupling nonlinear equation set:
wherein f isφ(x)、fQ(x)、
And fλ(x) Residual forms of additional equations, which are a steady-state neutron diffusion equation, a neutron flux-power equation, a coolant energy steady-state equation, a fuel heat conduction steady-state equation, and a closed neutron diffusion equation, respectively; x is a solution vector; phi is neutron flux; a isw、ae、an、as、at、ab、apRespectively are coefficients of all directions when the central mesh points are different; g is the total number of energy groups; χ is a neutron fission spectrum; sigmag′→gIs a diffusion cross section from the g 'th group to the g' th group; v sigmafCreating a cross-section for neutron fission; k is a radical ofeffEffective multiplication factor; Δ V is the grid volume; g is the g energy group; g 'is the g' th energy group; k sigmafIs an energy fission cross section; qi，j，kIs the actual power of the grid (i, j, k); t isfIs the fuel temperature; t iscIs the coolant temperature; q is the volumetric heat release rate; r is the fuel rod radius; h is the convective heat transfer coefficient of the coolant; lambda is the heat conductivity coefficient of the fuel rod; w is the coolant mass flow rate; cp is the constant pressure specific heat capacity of the coolant;
flux-power normalization coefficients; i, j and k are grid coordinates in x, y and z directions respectively;
and step S2, solving the nonlinear equation set by adopting a JFNK method to obtain the steady-state neutron flux, power, coolant temperature and fuel rod temperature of the whole reactor core fuel grid cell level.
2. The reactor steady-state physical thermal full-coupling fine numerical simulation method according to claim 1, wherein the steady-state neutron diffusion equation is spatially dispersed in a central point difference format, and the discrete expression of the g-th group neutron diffusion equation is obtained as follows:
wherein:
ap，g＝∑R，g，i，j，kΔVi，j，k-aw，g-ae，g-an，g-as，g-ab，g-at，g
wherein D is the diffusion coefficient; sigmaRTo remove cross-sections; Δ x is the x-direction grid spacing; Δ y is the y-direction grid spacing; Δ z is the z-direction grid spacing.
3. The reactor steady-state physical thermal full-coupling fine numerical simulation method of claim 1, wherein the steady-state neutron flux-power equation is spatially dispersed in a central point difference format, and a discrete expression of the neutron flux-power equation is obtained by:
4. the reactor steady-state physical thermal fully-coupled fine numerical simulation method of claim 1, wherein the coolant energy steady-state equation is discretized spatially by a finite volume method by using a single-channel model, and the discrete expression of the coolant energy steady-state equation is obtained by:
5. the reactor steady-state physical thermal full-coupling fine numerical simulation method of claim 1, wherein the discrete expression of the fuel heat conduction steady-state equation is as follows:
6. the reactor steady-state physical thermal full-coupling fine numerical simulation method of claim 1, wherein the expression of the additional equation of the closed neutron diffusion equation is as follows:
7. the reactor steady-state physical thermal full-coupling fine numerical simulation method of claim 1, wherein the method step S2 specifically comprises:
converting the system of nonlinear equations into a linear expression:
J(xk)δxk＝-F(xk)
wherein J is the attic ratio matrix of f (x) 0; δ x is the argument increment; x is the number ofkIs an independent variable;
solving the linear expression of the nonlinear equation set by adopting a JFNK method, wherein the specific solving process is as follows:
step S21, constructing an Attic ratio vector or an approximate vector of the Attic ratio vector; the Acigure ratio vector is the product of an Acigure ratio matrix and a Krylov subspace basis vector;
step S22, solving a nonlinear equation set J (x) by adopting a generalized minimum residual error method GMRESk)δxk＝-F(xk) Obtaining the independent variable increment delta xk；
Step S23, updating the argument xk+1＝xk+δxkCalculating environmental parameters calculated by the independent variables;
step S24, repeating the steps S21-S23 until | | | F (x)k+1)||≤eps，||F(xk+1) And | | is a second-order norm of the residual error of the nonlinear equation set, and eps is convergence precision.
8. A reactor steady-state physical thermal full-coupling fine numerical simulation system is characterized by comprising a model building module, an analysis module and an output module;
the model building module is used for building a reactor steady-state physical thermal coupling nonlinear equation system:
wherein f isφ(x)、fQ(x)、
And fλ(x) Residual forms of additional equations, which are a steady-state neutron diffusion equation, a neutron flux-power equation, a coolant energy steady-state equation, a fuel heat conduction steady-state equation, and a closed neutron diffusion equation, respectively; x is a solution vector; phi is neutron flux; a isw、ae、an、as、at、ab、apRespectively are coefficients of all directions when the central mesh points are different; g is the total number of energy groups; χ is a neutron fission spectrum; sigmag′→gIs a diffusion cross section from the g-th group to the g-th group; v sigmafCreating a cross-section for neutron fission; k is a radical ofeffEffective multiplication factor; Δ V is the grid volume; g is the g energy group; g 'is the g' th energy group; k sigmafIs an energy fission cross section; qi，j，kIs the actual power of the grid (i, j, k); t isfIs the fuel temperature; t iscIs the coolant temperature; q is the volumetric heat release rate; r is the fuel rod radius; h is the convective heat transfer coefficient of the coolant; lambda is the heat conductivity coefficient of the fuel rod; w is the coolant mass flow rate; cp is the constant pressure specific heat capacity of the coolant;
flux-power normalization coefficients; i, j and k are grid coordinates in x, y and z directions respectively;
the analysis module solves the nonlinear equation set by adopting a JFNK method to obtain the steady-state neutron flux, power, coolant temperature and fuel rod temperature of the fuel grid cell level of the whole reactor core;
and the output module is used for outputting the three-dimensional steady-state physical parameters and the thermal parameters of the whole reactor core fuel grid cell level obtained by the solution of the analysis module.
9. A computer device comprising a memory and a processor, the memory storing a computer program, characterized in that the processor, when executing the computer program, implements the steps of the method according to any of claims 1-7.
10. A computer-readable storage medium, on which a computer program is stored which, when being executed by a processor, carries out the steps of the method according to any one of claims 1 to 7.
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