A Fast Fuel Management Model for Liquid Metal Fast Reactors and Its Construction Method and Application

Abstract

A fast fuel management model for a liquid metal fast reactor, a method for constructing the same, and an application thereof, belonging to the technical field of nuclear reactor physics. To achieve fast three-dimensional core calculations, the present invention establishes a two-dimensional single assembly model, a super-assembly model, and a radial reflector model by constructing an entity geometry method; performs assembly calculations using the Monte Carlo continuous energy point cross-section neutron transport method; for core calculations, uses the nodal diffusion method to perform core calculations and solve the core neutron diffusion equation; combines an independent burnup calculation module to construct a core diffusion and burnup calculation model, accurately solve the burnup equation, and update the nuclear density; updates the macroscopic cross-section, uses the homogenized microscopic cross-section of the assembly at the initial moment and the nuclear density updated by the assembly burnup calculation to assemble the homogenized macroscopic cross-section of the assembly at the next burnup step and perform the core calculation of the next burnup step until the end of the core life cycle. The present invention realizes fast core burnup calculations and fast fuel management.

Description

Technical Field

[0001] The present invention belongs to the technical field of nuclear reactor physics, and particularly relates to a rapid fuel management model for a liquid metal fast reactor, a construction method thereof, and an application thereof. Background Art

[0002] Neutronics calculation and analysis is the core content of reactor core design and research and development. Usually, the Monte Carlo method or the deterministic method is used to solve the neutron diffusion equation or the transport equation. Currently, the one-step method or the two-step method can be used for neutronics analysis of fast reactors. The one-step method directly solves the three-dimensional full reactor, introduces fewer approximations, has accurate modeling, and high calculation accuracy. However, the cost of three-dimensional reactor core calculation is huge. In the neutronics calculation of fast reactors (especially nuclear design calculations), the two-step calculation process similar to that of pressurized water reactors is still widely used, that is, first obtain the homogenized few-group constants of each component region, and then use the reactor core program to perform three-dimensional reactor core neutronics calculation.

[0003] Due to the strong elastic scattering resonance phenomenon of medium-mass nuclides such as sodium and iron in the fast neutron region, the resonance effect in fast reactors is more complex than that in pressurized water reactors, and a finer energy group structure is required to analyze the energy self-shielding effect. Moreover, fast reactors have the characteristic of a hard energy spectrum and no longer have the "1 / E" spectrum characteristic. The neutronics analysis of fast reactors cannot directly use the two-step calculation model of pressurized water reactors. Therefore, it is necessary to specifically develop a two-step calculation model suitable for neutronics calculation and analysis of fast reactors. Moreover, at present, the domestic and foreign research on the two-step method of fast reactors based on the Monte Carlo and reactor core nodal methods mainly focuses on the calculation and analysis of the initial steady-state conditions, and there is less research on the microscopic burnup calculation method. Summary of the Invention

[0004] The problem to be solved by the present invention is to achieve rapid three-dimensional reactor core calculation, and propose a rapid fuel management model for a liquid metal fast reactor, a construction method thereof, and an application thereof.

[0005] To achieve the above object, the present invention is realized through the following technical solutions:

[0006] A construction method of a rapid fuel management model for a liquid metal fast reactor includes the following steps:

[0007] S1. Construct a component model, and establish a two-dimensional single-component model, a super-component model, and a radial reflector model by constructing an entity geometry method;

[0008] S2. Component calculation, and perform component calculation by using the Monte Carlo continuous energy point cross-section neutron transport method;

[0009] S3. Reactor core calculation, and solve the reactor core neutron diffusion equation by using the nodal diffusion method;

[0010] S4. Burnup calculation. Based on steps S2 and S3, combined with an independent burnup calculation module, a core diffusion and burnup calculation model is constructed to accurately solve the burnup equation and update the nuclear density;

[0011] S5. Update the macroscopic cross-section. Using the homogenized microscopic cross-section of the assembly at the initial moment and the nuclear density updated by the assembly burnup calculation, assemble the homogenized macroscopic cross-section of the assembly for the next burnup step, and then perform the core calculation for the next burnup step until the end of the core life cycle.

[0012] Furthermore, the specific implementation method of step S1 includes the following steps:

[0013] S1.1. Establish a two-dimensional single-assembly model through the CSG method. The two-dimensional single-assembly model applies a total reflection boundary and is established according to the actual fuel assembly layout.

[0014] S1.2. Establish a super-assembly model through the CSG method. The super-assembly model applies a total reflection boundary and is established by surrounding a non-breeding region assembly with a ring of fuel assemblies. The non-breeding region includes one of a bottom structure, an axial reflector, a gas chamber, a control assembly empty channel, a control assembly absorber, and a shielding layer.

[0015] S1.3. Establish a radial reflector model through the CSG method. The radial reflector model is a radial reflector model with fuel assemblies in the middle and radial reflector assemblies on the outside, generating the homogenized group constants for the radial reflector region. The left and right sides of the radial reflector model adopt vacuum boundaries, and the upper and lower sides adopt reflection boundaries.

[0016] Furthermore, the assembly in step S1 belongs to one of a lead-based traveling wave reactor, a lead-cooled fast reactor, a sodium-cooled fast reactor, and a lead-bismuth fast reactor.

[0017] Furthermore, the specific implementation method of step S2 includes the following steps:

[0018] S2.1. Use the Monte Carlo neutron transport calculation program OpenMC to perform neutron transport calculations based on the continuous energy point cross-sections in the ENDF / B-VII.1 evaluation nuclear database, and generate the homogenized group constants for the fuel assembly region and each non-breeding region through the assembly model constructed in step S1.

[0019] First, let the volume V of the homogenized region n n be composed of the volumes V of each calculation region m m The expression is:

[0020]

[0021] Among them, Σ is the summation operator. At the same time, the coarse energy group is composed of fine energy groups. To generate the multigroup cross section, merging is performed over the spatial region and energy groups. During the homogenization and merging process of energy groups and regions, the reaction rate before and after merging is ensured to be conserved. Then, there is:

[0022]

[0023] Among them, g is the coarse energy group number, e is the fine energy group number, t represents the total reaction, f represents the fission reaction, s represents the scattering reaction, g' represents the coarse energy group number before neutron reaction, e' represents the fine energy group number before neutron reaction, ν represents the average number of neutrons generated per fission, Σ t,e,m is the average macroscopic total cross section of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, νΣ f,e,m is the average macroscopic effective fission neutron production cross section of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, Σ s,e'→e,m is the average macroscopic scattering cross section from the e'-th fine energy group to the e-th fine energy group in the calculation region m during spatial homogenization and group merging, is the average macroscopic total cross section of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, is the average macroscopic effective fission neutron production cross section of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, is the average macroscopic scattering cross section from the g'-th coarse energy group to the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, φ e,m is the average neutron flux density of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, is the average neutron flux density of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging;

[0024] S2.2. Set the average flux after homogenization to be calculated from the flux before homogenization. The average cross section after homogenization is expressed as:

[0025]

[0026] Based on the conservation of the reaction rate before and after homogenization, the fission neutron energy spectrum calculation formula is as follows:

[0027]

[0028] S2.3. For the supercomponent model, based on the influence of materials and the influence of neutron flux distribution, the homogenization calculation process is corrected using the superhomogenization method SPH: Calculate using a two-loop supercomponent model, and iterate the SPH factor until convergence;

[0029] S2.3.1. Use the Monte Carlo neutron transport calculation program OpenMC to calculate the non-uniform two-loop assembly model, solve the neutron transport equation, and obtain the average neutron flux density and homogenized macroscopic cross-section results for energy group g and homogenized region n of the assembly. Fill in the homogenized few-group cross-section and fission neutron energy spectrum data into the diffusion program to construct the multi-group neutron diffusion equation:

[0030]

[0031] where r is the spatial vector, is the gradient operator, is the divergence operator, n is the homogenized region number, g is the coarse energy group number, G is the maximum coarse energy group number, t represents the total reaction, f represents the fission reaction, s represents the scattering reaction, g' represents the coarse energy group number before the neutron reaction, ν represents the average number of neutrons produced per fission, k represents the effective multiplication factor, MC represents the parameter calculated by the Monte Carlo program, χ g is the neutron fraction entering energy group g, φ g (r) is the neutron flux density of energy group g at position r, is the diffusion coefficient of energy group g at homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic total cross-section of energy group g at homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic fission cross-section of energy group g at homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic scattering cross-section from energy group g' to energy group g at homogenized region n calculated by the Monte Carlo program OpenMC;

[0032] S2.3.2. Use the core nodal diffusion program to calculate the equivalent uniform two-loop assembly model to solve the multi-group neutron diffusion equation, and obtain the average flux of the homogenized energy group g and nodal region n Then use the following formula to calculate the SPH factor:

[0033]

[0034] where Σ is the summation operator, n is the homogenized region number, g is the coarse energy group number, norm represents the normalized parameter, MC represents the parameter calculated by the Monte Carlo program, NODAL represents the parameter calculated by the core nodal diffusion program, represents the normalized SPH factor of homogenized region n and energy group g, represents the average flux of the homogenized region n of the assembly and energy group g calculated by the Monte Carlo program, represents the average flux of the homogenized region n of the assembly and the g-th energy group calculated by the core diffusion program, N g is the normalization factor, V n represents the volume of the homogenized region n;

[0035] Use the calculated SPH factor to correct the macroscopic cross-section:

[0036]

[0037] where x represents a specific reaction type, n is the homogenized region number, g is the coarse energy group number, and * represents the parameter after correction, represents the normalized SPH factor of the homogenized region n and the g-th energy group, Σ x,g,n represents the macroscopic cross-section of a specific reaction of the homogenized region n and the g-th energy group, represents the macroscopic cross-section of a specific reaction of the homogenized region n and the g-th energy group after SPH correction;

[0038] S2.3.3. Based on the SPH factor obtained in step S2.3.2 Correct the homogenized few-group constants of each energy group and each region, fill the corrected group constants into the diffusion program, and repeat steps S2.3.2 and S2.3.3 until the given convergence condition is met to obtain the final SPH factor and the homogenized few-group constants after SPH correction.

[0039] Furthermore, step S3 uses the homogenized macroscopic cross-sections of each component region to perform core nodal diffusion calculations, solve the three-dimensional full-core neutron diffusion equation, and obtain the core effective multiplication factor, power distribution, and neutron flux distribution results;

[0040] S3.1. The core nodal diffusion program is used to solve quadrilateral and hexagonal nodes, and the solution of the three-dimensional neutron diffusion equation is transformed into the solution of three one-dimensional transverse integral equations. Among them, the nodal expansion method NEM is mainly used to solve the quadrilateral geometry node, and the equations are coupled through the transverse leakage term;

[0041] S3.2. For the hexagonal geometry node, the neutron diffusion equation is solved using NEM in the axial direction. In the radial direction, the hexagonal node is divided into six triangular nodes, a local coordinate is set for each triangular node, and the triangular polynomial expansion nodal method TPEN is used for solution.

[0042] Furthermore, the specific implementation method of step S4 includes the following steps:

[0043] S4.1. In the initial state, based on the Monte Carlo program OpenMC, track the transmutation reactions in the burnup chain file to generate the microscopic parameter library required for independent burnup calculations. The transmutation reactions include fission, (n,γ), (n,2n), (n,3n), (n,4n), (n,p), and (n,α).

[0044] S4.2. Use the microscopic parameter library, burnup chain file, and the power distribution results obtained from the core diffusion calculation at the current burnup step to perform the assembly independent burnup calculation. OpenMC uses the Chebyshev rational approximation method to solve the matrix exponential, solve the Bateman equation, and update the nucleon density.

[0045] Furthermore, the specific process of updating the homogenized macroscopic cross-section at the burnup time point in step S5 is as follows:

[0046] Based on the condition of unchanged microscopic cross-section, use the nucleon density of the assembly at the next burnup moment and the initial homogenized microscopic cross-section of the assembly to assemble a new homogenized macroscopic cross-section of the assembly, or use the nucleon density of the assembly at the next burnup moment and the homogenized microscopic cross-section to assemble a new macroscopic cross-section. The formula for macroscopic cross-section assembly is:

[0047]

[0048] where x represents a specific reaction type, n is the homogenized region number, g is the coarse energy group number, i represents the nuclide, and Σ x,g,n represents the macroscopic cross-section of the specific reaction in the homogenized region n and the g-th energy group, and N i,n represents the nucleon density of nuclide i in the homogenized region n.

[0049] After obtaining the new homogenized macroscopic cross-section of the assembly, perform the core nodal diffusion calculation for the next burnup step.

[0050] A fast fuel management model for a liquid metal fast reactor is realized based on the construction method of the fast fuel management model for a liquid metal fast reactor. The fast fuel management model for a liquid metal fast reactor assembles the homogenized macroscopic cross-section of the assembly for the next burnup step using the initial homogenized microscopic cross-section of the assembly and the nucleon density updated by the assembly burnup calculation.

[0051] An application of the fast fuel management model for a liquid metal fast reactor is used in nuclear reactors, including lead-based traveling wave reactors, lead-cooled fast reactors, sodium-cooled fast reactors, and lead-bismuth fast reactors.

[0052] Advantages of the present invention:

[0053] A method for constructing a fast fuel management model for a liquid metal fast reactor according to the present invention. The method uses the Monte Carlo method to solve the neutron transport equation to calculate the homogenized group constants of components, and uses a more refined continuous energy neutron cross-section to analyze the energy self-shielding effect, which can accurately simulate the resonance interference phenomenon in the fast reactor, generate homogenized group constants with higher accuracy, and has strong geometric adaptability based on constructive solid geometry modeling. The calculated homogenized group constants of components are provided to the core nodal diffusion program to solve the neutron diffusion equation, realizing fast three-dimensional core calculation and obtaining results such as the core effective multiplication factor and power distribution.

[0054] A method for constructing a fast fuel management model for a liquid metal fast reactor according to the present invention performs component-independent burnup calculations using a pre-generated microscopic parameter library for burnup calculations and results such as the component power distribution obtained from core diffusion calculations at each burnup step, updates the nucleon density, realizes fast core burnup calculations, and ultimately realizes fast fuel management.

[0055] A method for constructing a fast fuel management model for a liquid metal fast reactor according to the present invention can assemble the homogenized macroscopic cross-section of the next burnup step using the homogenized microscopic cross-section of the component at the initial moment and the nucleon density updated by component burnup calculations, without pre-generating a large number of homogenized macroscopic cross-sections at different burnup times for interpolation calculations. BRIEF DESCRIPTION OF THE DRAWINGS

[0056] Figure 1 is a flow chart of a method for constructing a fast fuel management model for a liquid metal fast reactor according to the present invention;

[0057] Figure 2 is a detailed flow chart of fast burnup calculation for a lead-cooled traveling wave reactor in Embodiment 1 of the present invention;

[0058] Figure 3 are a two-dimensional single-component model and a super-component model for Monte Carlo component homogenization parameter calculation in Embodiment 1 of the present invention; where (a) is a two-dimensional single-component model for fuel component homogenization parameter calculation, and (b)-(h) are two-dimensional super-component models for non-fission region homogenization parameter calculation. (b) is a super-component model for the empty channel region, (c) is a super-component model for the absorber region, (d) is a super-component model for the shielding region, (e) is a super-component model for the coolant region, (f) is a super-component model for the bottom structure region, (g) is a super-component model for the axial reflector region, and (h) is a super-component model for the gas chamber region;

[0059] Figure 4 is a radial reflector model for Monte Carlo component homogenization calculation in Embodiment 1 of the present invention;

[0060] Figure 5The component model for calculating the super homogenization (SPH) factor in Embodiment 1 of the present invention, where (a) is a non-uniform two-loop super-component model and (b) is an equivalent uniform two-loop super-component model;

[0061] Figure 6 The iterative calculation flow chart for obtaining the SPH factor in Embodiment 1 of the present invention;

[0062] Figure 7 The deviation of the core effective multiplication factor relative to the reference solution obtained by using the microscopic cross section varying with burnup to assemble the macroscopic cross section and performing burnup calculation in Embodiment 1 of the present invention;

[0063] Figure 8 The deviation of the nucleon density of the main nuclides relative to the reference solution in Embodiment 1 of the present invention;

[0064] Figure 9 The deviation of the core effective multiplication factor relative to the reference solution obtained by using the microscopic cross section varying with burnup and the nucleon density at the corresponding burnup step of the reference solution to assemble the macroscopic cross section in Embodiment 1 of the present invention. Detailed implementation manners

[0065] In order to make the purpose, technical solutions and advantages of the present invention clearer, the present invention will be further described in detail below with reference to the accompanying drawings and specific implementation manners. It should be understood that the specific implementation manners described herein are only used to explain the present invention and are not used to limit the present invention, that is, the described specific implementation manners are only a part of the implementation manners of the present invention, rather than all the specific implementation manners. Usually, the components of the specific implementation manners of the present invention described and shown in the accompanying drawings herein can be arranged and designed in various different configurations, and the present invention can also have other implementation manners.

[0066] Therefore, the detailed description of the specific implementation manners of the present invention provided in the accompanying drawings below is not intended to limit the scope of the claimed present invention, but only represents the selected specific implementation manners of the present invention. Based on the specific implementation manners of the present invention, all other specific implementation manners obtained by those skilled in the art without creative efforts belong to the scope of protection of the present invention.

[0067] To further understand the content, features and effects of the present invention, the following specific implementation manners are exemplified and are accompanied by the Figure 1 -Accompanying Figure 9 The details are as follows:

[0068] Embodiment 1:

[0069] A method for constructing a fast fuel management model for a liquid metal fast reactor, comprising the following steps:

[0070] S1. Build component models, and establish two-dimensional single-component models, super-component models, and radial reflector models by constructing entity geometries.

[0071] Further, the specific implementation method of step S1 includes the following steps:

[0072] S1.1. Establish a two-dimensional single-component model by the CSG method. The two-dimensional single-component model applies a total reflection boundary and is established according to the actual fuel assembly layout.

[0073] S1.2. Establish a super-component model by the CSG method. The super-component model applies a total reflection boundary and is established by surrounding a non-fissile region component with a ring of fuel assemblies. The non-fissile region includes one of a bottom structure, an axial reflector, a gas chamber, a control assembly empty channel, a control assembly absorber, and a shielding layer.

[0074] S1.3. Establish a radial reflector model by the CSG method. The radial reflector model is a radial reflector model with fuel assemblies in the middle and radial reflector assemblies on the outside, generates homogenized group constants for the radial reflector region, and uses vacuum boundaries on the left and right sides and reflection boundaries on the top and bottom sides of the radial reflector model.

[0075] Further, the component in step S1 belongs to one of a lead-based traveling wave reactor, a lead-cooled fast reactor, a sodium-cooled fast reactor, and a lead-bismuth fast reactor.

[0076] Further, the fuel assembly can not only provide neutrons for the calculation of non-fissile region homogenized group constants but also simulate the actual core environment.

[0077] This embodiment provides a method for constructing a fast fuel management model for a lead-based traveling wave reactor facing a liquid metal fast reactor. The lead-based traveling wave reactor (standing wave reactor) adopts the core layout of a typical sodium-cooled fast reactor and realizes long-term reactor breeding combustion through a suitable refueling scheme.

[0078] To handle the complex resonance calculation problem in the fast neutron energy region, the continuous energy point cross-section data required for solving the neutron transport equation by the Monte Carlo method is given by the currently publicly available evaluated nuclear database.

[0079] S2. Component calculation, using the Monte Carlo continuous energy point cross-section neutron transport method for component calculation; in the process of generating homogenized group constants in component calculation, the core is homogenization and energy group merging.

[0080] Further, the specific implementation method of step S2 includes the following steps:

[0081] S2.1. Use the Monte Carlo neutron transport calculation program OpenMC to perform neutron transport calculations based on the continuous energy point cross-sections in the ENDF / B-VII.1 evaluated nuclear database, and generate the homogenized group constants for the fuel assembly region and each non-breeding region through the component model constructed in step S1;

[0082] First, let the volume V of the homogenized region n n be composed of the volumes V m of each calculation region m, and the expression is:

[0083]

[0084] where Σ is the summation operator. At the same time, let the coarse energy group be composed of the fine energy groups. To generate the multi-group cross-sections, merge in the spatial region and energy group; during the homogenization and merging process of the energy group and region, ensure that the reaction rate is conserved before and after merging, then there is:

[0085]

[0086] where g is the coarse energy group number, e is the fine energy group number, t represents the total reaction, f represents the fission reaction, s represents the scattering reaction, g' represents the coarse energy group number before the neutron reaction, e' represents the fine energy group number before the neutron reaction, ν represents the average number of neutrons generated per fission, Σ t,e,m is the average macroscopic total cross-section of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, νΣ f,e,m is the average macroscopic effective fission neutron production cross-section of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, Σ s,e'→e,m is the average macroscopic scattering cross-section of the calculation region m from the e'-th fine energy group to the e-th fine energy group during spatial homogenization and group merging, is the average macroscopic total cross-section of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, is the average macroscopic effective fission neutron production cross-section of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, is the average macroscopic scattering cross-section of the homogenized region n from the g'-th coarse energy group to the g-th coarse energy group after spatial homogenization and group merging, φ e,m is the average neutron flux density of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, is the average neutron flux density of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging;

[0087] S2.2. Set the average flux after homogenization to be obtained from the flux before homogenization, and the average cross-section after homogenization is expressed as:

[0088]

[0089] When the spatially homogenized merged region is a single region, such an equivalence does not introduce an approximation. However, for the problem of merging into two regions (supercomponent model), especially in the case where the material differences are very large, such as calculating the supercomponent model of the absorber region ( Figure 3 (c)), due to the presence of a strong absorption region, there are significant differences in the neutron flux distribution, and there is a certain approximation in the above homogenization calculation process, and the SPH method needs to be used for correction. For the fission neutron energy spectrum, the actual homogenization calculation does not introduce an approximation, and the reaction rate is conserved before and after homogenization, so SPH correction is not required.

[0090] Based on the conservation of the reaction rate before and after homogenization, the calculation formula for the fission neutron energy spectrum is as follows:

[0091]

[0092] The key to SPH correction is the calculation of the SPH factor. Due to geometric modeling limitations in some core nodal diffusion programs, it is impossible to equivalently calculate supercomponent models such as Figure 3 (b)-(h), so a two-loop supercomponent model ( Figure 5 ) is used for calculation. The SPH factor usually needs to be iteratively calculated until convergence, as shown in Figure 6 .

[0093] S2.3. For the supercomponent model, based on the influence of materials and neutron flux distribution, the homogenization calculation process is corrected using the super homogenization method SPH: calculate using a two-loop supercomponent model, and iteratively calculate the SPH factor until convergence;

[0094] S2.3.1. Use the Monte Carlo neutron transport calculation program OpenMC to calculate the heterogeneous two-loop component model, solve the neutron transport equation, obtain the average neutron flux density and homogenized macroscopic cross-section results for energy group g and homogenized region n of the component, fill in the homogenized few-group cross-section and fission neutron energy spectrum data into the diffusion program, and construct a multi-group neutron diffusion equation:

[0095]

[0096] Among them, r is the spatial vector, is the gradient operator, is the divergence operator, n is the homogenized region number, g is the coarse energy group number, G is the maximum coarse energy group number, t represents the total reaction, f represents the fission reaction, s represents the scattering reaction, g' represents the coarse energy group number before the neutron reaction, ν represents the average number of neutrons produced per fission, k represents the effective multiplication factor, MC represents the parameters calculated by the Monte Carlo program, χ g is the neutron fraction entering energy group g, and φ g (r) is the neutron flux density of energy group g at position r. is the diffusion coefficient of the g-th energy group at the homogenized region n calculated by the Monte Carlo code OpenMC, is the average macroscopic total cross-section of the g-th energy group at the homogenized region n calculated by the Monte Carlo code OpenMC, is the average macroscopic fission cross-section of the g-th energy group at the homogenized region n calculated by the Monte Carlo code OpenMC, is the average macroscopic scattering cross-section from the g'-th energy group to the g-th energy group at the homogenized region n calculated by the Monte Carlo code OpenMC;

[0097] S2.3.2. Solve the multi-group neutron diffusion equation using the core nodal diffusion code to calculate the equivalent homogeneous two-loop assembly model, and obtain the average flux of the homogenized energy group g and nodal region n Then calculate the SPH factor using the following formula:

[0098]

[0099] where Σ is the summation operator, n is the homogenized region number, g is the energy group number, norm represents the normalized parameter, MC represents the parameter calculated by the Monte Carlo code, and NODAL represents the parameter calculated by the core nodal diffusion code, represents the normalized SPH factor of the homogenized region n and the g-th energy group, represents the average flux of the homogenized region n and the g-th energy group of the assembly calculated by the Monte Carlo code, represents the average flux of the homogenized region n and the g-th energy group of the assembly calculated by the core diffusion code, N g is the normalization factor, V n represents the volume of the homogenized region n;

[0100] Correct the macroscopic cross-section using the calculated SPH factor:

[0101]

[0102] where x represents a specific reaction type, n is the homogenized region number, g is the energy group number, and * represents the corrected parameter, represents the normalized SPH factor of the homogenized region n and the g-th energy group, Σ x,g,n represents the macroscopic cross-section of the specific reaction of the homogenized region n and the g-th energy group, represents the macroscopic cross-section of the specific reaction of the homogenized region n and the g-th energy group after SPH correction;

[0103] S2.3.3. Based on the SPH factor obtained in step S2.3.2 Modify the homogenized few-group constants for each energy group and each region, fill the modified group constants into the diffusion program, and repeat steps S2.3.2 and S2.3.3 until the given convergence condition is met to obtain the final SPH factor and the homogenized few-group constants corrected by SPH.

[0104] S3. Core calculation: Solve the core neutron diffusion equation using the nodal diffusion method.

[0105] Furthermore, in step S3, use the homogenized macroscopic cross-sections of each component region to perform core nodal diffusion calculations, solve the three-dimensional full-core neutron diffusion equation, and obtain the results of the core effective multiplication factor, power distribution, and neutron flux distribution.

[0106] S3.1. The core nodal diffusion program is used to solve quadrilateral and hexagonal nodes, transforming the solution of the three-dimensional neutron diffusion equation into the solution of three one-dimensional transverse integral equations. Among them, the nodal expansion method NEM is mainly used to solve the quadrilateral geometry nodes, and the equations are coupled through the transverse leakage term.

[0107] S3.2. For hexagonal geometry nodes, use NEM to solve the neutron diffusion equation axially. Radially, divide the hexagonal node into six triangular nodes, set a local coordinate for each triangular node, and use the triangular polynomial expansion nodal method TPEN to solve.

[0108] Furthermore, the main idea of the TPEN method is to use the nodal expansion method (NEM) to solve the neutron diffusion equation axially, divide the hexagonal node into six triangular nodes radially, set a local coordinate for each triangular node, and take the outer normal direction of each hexagonal surface as the positive x direction of the local coordinate. The transverse integral neutron diffusion equation within a hexagonal node region consists of six radial equations and one axial equation, and the form of the neutron balance equation set is as follows:

[0109]

[0110] Among them, the subscript g represents the g-th coarse energy group, and m represents the m-th triangle.

[0111] The lateral leakage source terms represented in the system of equations have a spatial dependence, which in the TPEN method is determined by the average of the nodal lateral leakage of the neighboring nodes and the current node itself. At the beginning of the iterative calculation, all lateral leakages are considered to be known, and the corresponding two equilibrium equations are independent and can be solved separately. Afterwards, the lateral leakage of the current node is considered to be unknown, and the radial and axial equations are coupled through the lateral leakage terms. The source terms are approximated by seven two-dimensional polynomials, and the seven coefficients are determined by retaining the nodal average of the source terms at the six surrounding nodes and the node of interest. The radial equations in the system of equations are solved by the following nine-term two-variable polynomial representing the flux within the triangular node:

[0112]

[0113] By combining all equations and relations, adding constraints and processing the equations appropriately, we can obtain a complete iterative solution formula. By selecting a suitable numerical iteration method, we can obtain the core neutronics results.

[0114] S4. Burnup calculation: Based on step S2 and step S3, an independent burnup calculation module is combined to construct a core diffusion and burnup calculation model, accurately solve the burnup equation, and update the nuclear density;

[0115] Furthermore, the specific implementation method of step S4 includes the following steps:

[0116] S4.1. In the initial state, the transmutation reaction in the burnup chain file is tracked based on the Monte Carlo program OpenMC to generate a microscopic parameter library required for independent burnup calculation, wherein the transmutation reaction includes fission, (n,γ), (n,2n), (n,3n), (n,4n), (n,p) and (n,α);

[0117] S4.2. Component independent burnup calculations are performed using the microscopic parameter library, burnup chain file, and the power distribution results obtained from the core diffusion calculations at the current burnup step. OpenMC uses the Chebyshev rational approximation method to solve the matrix index, solve the Bateman equation, and update the nucleon density.

[0118] S5. Update the macroscopic cross section, use the component homogenized microscopic cross section and component burnup calculation at the initial moment to update the nuclear density, assemble the component homogenized macroscopic cross section for the next burnup step, and then perform the core calculation for the next burnup step until the end of the core life.

[0119] Furthermore, the specific process of updating the homogenized macroscopic cross section with the burnup time point in step S5 is:

[0120] Based on the condition of invariant microscopic cross-section, a new homogenized macroscopic cross-section of the assembly is obtained by assembling the nuclide density of the assembly at the next burnup moment and the homogenized microscopic cross-section at the initial moment, or a new macroscopic cross-section is obtained by assembling the nuclide density of the assembly at the next burnup moment and the homogenized microscopic cross-section; the formula for macroscopic cross-section assembly is:

[0121]

[0122] where x represents a specific reaction type, n is the number of homogenized regions, g is the number of coarse energy groups, i represents a nuclide, and Σ x,g,n represents the macroscopic cross-section of a specific reaction for homogenized region n and the g-th energy group, and N i,n represents the nuclide density of nuclide i in homogenized region n;

[0123] After obtaining the new homogenized macroscopic cross-section of the assembly, perform the core nodal diffusion calculation for the next burnup step.

[0124] The homogenized macroscopic cross-section is an important homogenized parameter required for neutronics calculations of the reactor core, and its value is jointly determined by the microscopic cross-sections of various nuclides and the nuclide densities of various nuclides. Since the microscopic cross-sections and nuclide densities of nuclides usually change with the operation time (burnup time) of the reactor, the core calculation needs to use the homogenized macroscopic cross-section corresponding to the current burnup time point. In the macroscopic burnup scheme, a series of homogenized macroscopic cross-sections of the assembly at different burnup time points are usually calculated in advance and then the homogenized macroscopic cross-section at the required burnup time point is obtained by interpolation. In the microscopic burnup scheme, after each core calculation is completed, the nuclide densities of each nuclide at the next burnup time point in each burnup region of the core are obtained by performing burnup calculations using the core neutron flux distribution and power distribution calculation results at the current burnup time point. A new homogenized macroscopic cross-section is obtained by assembling the new nuclide density and the homogenized microscopic cross-section, and then the new homogenized macroscopic cross-section can be used for the next core calculation.

[0125] After obtaining the new homogenized macroscopic cross-section of the assembly, the core nodal diffusion calculation for the next burnup step can be performed. Since the core nodal diffusion method can efficiently solve the three-dimensional core neutron diffusion equation, its calculation efficiency is better than that of the Monte Carlo continuous energy point cross-section full-core transport calculation, and it can quickly solve the core effective multiplication factor, power distribution, neutron flux distribution, etc. results at each burnup step. Combining with the assembly independent burnup calculation can achieve fast core diffusion and burnup calculations, and then achieve fast fuel management.

[0126] Figure 7 This is the deviation of the core effective multiplication factor obtained by using the microscopic cross-section varying with burnup to assemble the macroscopic cross-section and performing burnup calculations in the embodiments of the present invention relative to the reference solution. The following briefly analyzes the errors in the calculation results of the embodiments of the present invention. First, analyze the nuclide densities of the main nuclides,Figure 8 This shows the deviation of the nucleon density of the main nuclides in the embodiments of the present invention relative to the reference solution. The calculation results show that the relative deviation of the nucleon density of most main nuclides is within 1%. Then, the influence of the calculation deviation of the nucleon density on the effective multiplication factor of the reactor core is analyzed. Figure 9 This shows the deviation of the effective multiplication factor of the reactor core obtained by assembling the macroscopic cross-section with the microscopic cross-section varying with burnup and the nucleon density corresponding to the burnup step of the reference solution in the embodiments of the present invention relative to the reference solution. The calculation results show that when the nucleon density is the same at each burnup step, the deviation of the effective multiplication factor of the reactor core calculated in the embodiments of the present invention relative to the reference solution remains near -100 pcm.

[0127] Embodiment 2:

[0128] A fast fuel management model for a liquid metal fast reactor is implemented based on the construction method of a fast fuel management model for a liquid metal fast reactor described in Embodiment 1. The fast fuel management model for a liquid metal fast reactor assembles the homogenized macroscopic cross-section of the next burnup step by using the homogenized microscopic cross-section of the components at the initial moment and the nucleon density calculated by component burnup update.

[0129] Furthermore, the nuclear reactor is a lead-based traveling wave reactor;

[0130] Furthermore, the lead-based traveling wave reactor uses lead as the coolant. The traveling wave reactor is also called a standing wave reactor, which belongs to a type of conceptual design of fast neutron reactors and can adopt the core layout of a typical sodium-cooled fast reactor.

[0131] Furthermore, the specific process for calculating the homogenized group constants in each component region of the lead-cooled traveling wave reactor is as follows: First, neutron transport calculations are performed using Monte Carlo programs such as OpenMC and Serpent based on the continuous energy point cross-sections in the evaluated nuclear database, and the homogenized group constants of the fuel region and the non-fission region of the lead-based traveling wave reactor are generated using two-dimensional single-component models and super-component models respectively. Due to the non-negligible energy spectrum shift effect at the interface between the driver component and the radial reflector, the homogenized group constants of the radial reflector region are calculated separately using a two-dimensional radial reflector model. And appropriate homogenization correction methods such as the super homogenization (SPH) method are used to correct the calculation results of the homogenized group constants in the strong absorption region to ensure the conservation of the neutron reaction rate in each homogenized region before and after homogenization.

[0132] Furthermore, the calculation process of core diffusion and burnup in a lead-cooled traveling wave reactor is as follows: First, using the homogenized group constants of each component region calculated at the initial time, perform the core nodal diffusion calculation at the initial time of the core to obtain results such as the effective multiplication factor and power distribution of the core. Then, provide the power distribution and neutron flux distribution results calculated for the core to an independent burnup calculation program to solve the Bateman equation and obtain the nucleon densities of each component at the next burnup time. Next, use the homogenized microscopic cross-sections at the initial time (or the microscopic cross-sections varying with burnup obtained by prior component burnup calculations) and assemble them with the nucleon densities of each component at the next burnup time obtained from independent burnup calculations to generate homogenized macroscopic cross-sections. After obtaining the new homogenized macroscopic cross-sections of the components, the core nodal diffusion calculation for the next burnup step can be performed, thereby realizing the core burnup calculation and fuel management.

[0133] Example 3:

[0134] The application of a fast fuel management model for a liquid metal fast reactor is realized relying on the construction method of a fast fuel management model for a liquid metal fast reactor described in Example 1, and is used for nuclear reactors, including lead-based traveling wave reactors, lead-cooled fast reactors, sodium-cooled fast reactors, and lead-bismuth fast reactors.

[0135] Example 4:

[0136] A fast fuel management model for a liquid metal fast reactor is realized relying on the construction method of a fast fuel management model for a liquid metal fast reactor described in Example 1. The fast fuel management model for a liquid metal fast reactor assembles the homogenized macroscopic cross-sections of the components at the next burnup step using the homogenized microscopic cross-sections of the components at the initial time and the nucleon densities updated by component burnup calculations.

[0137] Furthermore, the nuclear reactor is a lead-cooled fast reactor;

[0138] Furthermore, the lead-cooled fast reactor uses lead as the coolant and belongs to a type of fast neutron reactor.

[0139] Furthermore, the specific process for calculating the homogenized group constants of each component region in a lead-cooled fast reactor is as follows: First, use Monte Carlo programs such as OpenMC and Serpent to perform neutron transport calculations based on the continuous energy point cross-sections in the evaluated nuclear database, and generate the homogenized group constants of the fuel region and non-fission region of the lead-based traveling wave reactor using two-dimensional single-component models and super-component models respectively. Due to the non-negligible energy spectrum shift effect at the interface between the driver component and the radial reflector, the homogenized group constants of the radial reflector region are calculated separately using a two-dimensional radial reflector model. And use appropriate homogenization correction methods such as the super homogenization (SPH) method to correct the calculation results of the homogenized group constants in the strong absorption region to ensure the conservation of neutron reaction rates in each homogenized region before and after homogenization.

[0140] Furthermore, the calculation process of the core diffusion and burnup of the lead-cooled fast reactor is as follows: First, using the homogenized group constants of each component region calculated at the initial moment, perform the core nodal diffusion calculation at the initial moment of the core to obtain results such as the effective multiplication factor and power distribution of the core. Then, provide the power distribution and neutron flux distribution results calculated for the core to an independent burnup calculation program to solve the Bateman equation and obtain the nucleon densities of each component at the next burnup moment. Next, use the homogenized microscopic cross-sections at the initial moment (or the microscopic cross-sections that vary with burnup obtained by prior component burnup calculations), and assemble them with the nucleon densities of each component at the next burnup moment obtained from the independent burnup calculation to generate homogenized macroscopic cross-sections. After obtaining the new homogenized macroscopic cross-sections of the components, the core nodal diffusion calculation for the next burnup step can be performed, thereby realizing the core burnup calculation and fuel management.

[0141] Example 5:

[0142] A fast fuel management model for a liquid metal fast reactor is realized based on the construction method of a fast fuel management model for a liquid metal fast reactor described in Example 1. The fast fuel management model for a liquid metal fast reactor assembles the homogenized macroscopic cross-sections of the components for the next burnup step using the homogenized microscopic cross-sections of the components at the initial moment and the nucleon densities updated by the component burnup calculation.

[0143] Furthermore, the nuclear reactor is a sodium-cooled fast reactor;

[0144] Furthermore, the sodium-cooled fast reactor uses sodium as the coolant and belongs to a type of fast neutron reactor.

[0145] Furthermore, the specific process for calculating the homogenized group constants of each component region of the sodium-cooled fast reactor is as follows: First, use Monte Carlo programs such as OpenMC and Serpent to perform neutron transport calculations based on the continuous energy point cross-sections in the evaluated nuclear database, and generate the homogenized group constants of the fuel region and non-fission region of the lead-based traveling wave reactor using two-dimensional single-component models and super-component models respectively. Due to the non-negligible energy spectrum shift effect at the interface between the driver component and the radial reflector, the homogenized group constants of the radial reflector region are calculated separately using a two-dimensional radial reflector model. And use a suitable homogenization correction method such as the super homogenization (SPH) method to correct the calculation results of the homogenized group constants in the strong absorption region to ensure the conservation of the neutron reaction rate in each homogenized region before and after homogenization.

[0146] Furthermore, the calculation process of core diffusion and burnup in a sodium-cooled fast reactor is as follows: First, using the homogenized group constants of each component region calculated at the initial moment, perform the core nodal diffusion calculation at the initial moment of the core to obtain results such as the effective multiplication factor and power distribution of the core. Then, provide the power distribution and neutron flux distribution results calculated for the core to an independent burnup calculation program to solve the Bateman equation and obtain the nucleon density of each component at the next burnup moment. Next, use the homogenized microscopic cross-section at the initial moment (or the microscopic cross-section varying with burnup obtained by prior component burnup calculation) and the nucleon density of each component at the next burnup moment obtained from the independent burnup calculation to assemble and generate the homogenized macroscopic cross-section. After obtaining the new homogenized macroscopic cross-section of the component, the core nodal diffusion calculation for the next burnup step can be performed, thereby realizing the core burnup calculation and fuel management.

[0147] Example 6:

[0148] A fast fuel management model for a liquid metal fast reactor is realized based on the construction method of a fast fuel management model for a liquid metal fast reactor described in Example 1. The fast fuel management model for a liquid metal fast reactor assembles the homogenized macroscopic cross-section of the component for the next burnup step using the homogenized microscopic cross-section of the component at the initial moment and the nucleon density updated by component burnup calculation.

[0149] Furthermore, the nuclear reactor is a lead-bismuth fast reactor;

[0150] Furthermore, the lead-bismuth fast reactor uses lead-bismuth as the coolant and belongs to a type of fast neutron reactor.

[0151] Furthermore, the specific process for calculating the homogenized group constants of each component region in a lead-bismuth fast reactor is as follows: First, use Monte Carlo programs such as OpenMC and Serpent to perform neutron transport calculations based on the continuous energy point cross-sections in the evaluated nuclear database, and use two-dimensional single-component models and super-component models to generate the homogenized group constants of the fuel region and non-fissile region of a lead-based traveling wave reactor respectively. Due to the non-negligible energy spectrum shift effect at the interface between the driver component and the radial reflector, the homogenized group constants of the radial reflector region are calculated separately using a two-dimensional radial reflector model. And use a suitable homogenization correction method such as the super homogenization (SPH) method to correct the calculation results of the homogenized group constants in the strong absorption region to ensure the conservation of the neutron reaction rate in each homogenized region before and after homogenization.

[0152] Furthermore, the calculation process of core diffusion and burnup for the lead-bismuth fast reactor is as follows: First, using the homogenized group constants of each component region calculated at the initial time, perform the core nodal diffusion calculation at the initial time of the core to obtain results such as the effective multiplication factor and power distribution of the core. Then, provide the power distribution and neutron flux distribution results calculated for the core to an independent burnup calculation program to solve the Bateman equation and obtain the nucleon density of each component at the next burnup time. Next, use the homogenized microscopic cross-section at the initial time (or the microscopic cross-section varying with burnup obtained by prior component burnup calculation) and assemble it with the nucleon density of each component at the next burnup time obtained from the independent burnup calculation to generate the homogenized macroscopic cross-section. After obtaining the new homogenized macroscopic cross-section of the component, the core nodal diffusion calculation for the next burnup step can be performed, thereby realizing the core burnup calculation and fuel management.

[0153] It should be noted that relational terms such as "first" and "second" are only used to distinguish one entity or operation from another entity or operation, and do not necessarily require or imply any actual relationship or order between these entities or operations. Moreover, the terms "comprising", "including" or any other variant thereof are intended to cover non-exclusive inclusion, so that a process, method, article or device comprising a series of elements includes not only those elements but also other elements not expressly listed, or elements inherent to such process, method, article or device. Without further limitation, an element defined by the statement "comprising an..." does not exclude the existence of additional identical elements in the process, method, article or device comprising the element.

[0154] Although the present application has been described above with reference to specific embodiments, various improvements can be made to it and components can be replaced with equivalents without departing from the scope of the present application. In particular, as long as there is no structural conflict, the various features in the specific embodiments disclosed in the present application can be combined with each other in any way, and the exhaustive description of these combinations is not given in this specification only for the sake of saving space and resources. Therefore, the present application is not limited to the specific embodiments disclosed herein, but includes all technical solutions falling within the scope of the claims.

Claims

1. A method for constructing a fast fuel management model for a liquid metal fast reactor, characterized in that, It includes the following steps: S1. Construct a component model, and establish a two-dimensional single-component model, a super-component model, and a radial reflector model by constructing entity geometry methods; S2. Component calculation, and perform component calculation using the Monte Carlo continuous-energy point cross-section neutron transport method; The specific implementation method of step S2 includes the following steps: S2.

1. Use the Monte Carlo neutron transport calculation program OpenMC to perform neutron transport calculation based on the continuous-energy point cross-section in the ENDF / B-VII.1 evaluated nuclear database, and generate the homogenized group constants of the fuel assembly region and each non-fissile region through the component model constructed in step S1; First, let the volume V of the homogenization region n n be composed of the volumes V of the respective calculation regions m m and the expression is as follows: Where ∑ is the summation operator, and at the same time let the coarse energy group be composed of fine energy groups. To generate multi-group cross-sections, merging is performed in the spatial region and energy group; during the homogenization merging process of the energy group and region, ensure that the reaction rate is conserved before and after merging, then there is: Among them, g is the coarse energy group number, e is the fine energy group number, t represents the total reaction, f represents the fission reaction, s represents the scattering reaction, g' represents the coarse energy group number before the neutron reaction, e' represents the fine energy group number before the neutron reaction, ν represents the average number of neutrons produced per fission, ∑ t,e,m is the average macroscopic total cross-section of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, v∑ f,e,m is the average macroscopic effective fission neutron production cross-section of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, ∑ s,e'→e,m is the average macroscopic scattering cross-section from the e'-th fine energy group to the e-th fine energy group in the calculation region m after spatial homogenization and group merging, is the average macroscopic total cross-section of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, is the average macroscopic effective fission neutron production cross-section of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, is the average macroscopic scattering cross-section from the g'-th coarse energy group to the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging, φ e,m is the average neutron flux density of the e-th fine energy group in the calculation region m before spatial homogenization and group merging, is the average neutron flux density of the g-th coarse energy group in the homogenized region n after spatial homogenization and group merging; S2.

2. Set that the homogenized average flux is obtained by calculating the flux before homogenization, and the homogenized average cross-section is expressed as: Based on the conservation of the reaction rate before and after homogenization, the fission neutron energy spectrum calculation formula is as follows: S2.

3. For the super-component model, based on the influence of materials and neutron flux distribution, the homogenization calculation process is corrected using the super homogenization method SPH: calculate using a two-loop super-component model, and iterate the SPH factor until convergence; S2.3.

1. Use the Monte Carlo neutron transport calculation program OpenMC to calculate the heterogeneous two-loop component model, solve the neutron transport equation, obtain the average neutron flux density and homogenized macroscopic cross-section results of energy group g and component homogenized region n, fill the homogenized few-group cross-section and fission neutron energy spectrum data into the diffusion program, and construct a multi-group neutron diffusion equation: where r is a spatial vector, is the gradient operator, is the divergence operator, g is the coarse energy group number, G is the maximum coarse energy group number, t represents the total reaction, f represents the fission reaction, s represents the scattering reaction, g' represents the coarse energy group number before the neutron reaction, ν represents the average number of neutrons produced per fission, k represents the effective multiplication factor, MC represents the parameter calculated by the Monte Carlo code, χ g is the neutron fraction entering the g-th energy group, φ g (r) is the neutron flux density of the g-th energy group at position r, is the diffusion coefficient of the g-th energy group in homogenized region n calculated by the Monte Carlo code OpenMC, is the average macroscopic total cross section of the g-th energy group in homogenized region n calculated by the Monte Carlo code OpenMC, is the average macroscopic fission cross section of the g-th energy group in homogenized region n calculated by the Monte Carlo code OpenMC, is the average macroscopic scattering cross section from the g'-th energy group to the g-th energy group in homogenized region n calculated by the Monte Carlo code OpenMC; S2.3.

2. Use the core subassembly diffusion code to calculate the equivalent uniform two-loop assembly model to solve the multi-group neutron diffusion equation, and obtain the average flux of energy group g and homogenized region n after homogenization. Furthermore, use the following formula to calculate the SPH factor: Among them, Σ is the summation operator, g is the coarse energy group number, norm represents the normalized parameter, MC represents the parameter calculated by the Monte Carlo program, and NODAL represents the parameter calculated by the core nodal diffusion program. represents the normalized SPH factor for the homogenized region n and the g-th energy group. represents the average flux of the homogenized region n of the assembly and the g-th energy group calculated by the Monte Carlo program. represents the average flux of the homogenized region n of the assembly and the g-th energy group calculated by the core diffusion program, N g is the normalization factor, V n represents the volume of the homogenized region n. Use the calculated SPH factor to correct the macroscopic cross-section: where x represents a specific reaction type, g is the coarse energy group number, and * represents the parameter after being corrected. represents the normalized SPH factor of homogenized region n and the g-th energy group, Σ x,g,n represents the macroscopic cross-section of the specific reaction of homogenized region n and the g-th energy group. represents the macroscopic cross-section of the specific reaction of homogenized region n and the g-th energy group after SPH correction; S2.3.

3. Based on the SPH factors obtained in step S2.3.2 Correct the homogenized few-group constants in each energy group and each region, fill the corrected group constants into the diffusion program, and repeat steps S2.3.2 and S2.3.3 until the given convergence condition is met to obtain the final SPH factors and the homogenized few-group constants corrected by SPH; S3. Core calculation, and perform core calculation using the nodal diffusion method to solve the core neutron diffusion equation; S4. Burnup calculation, based on steps S2 and S3, combined with an independent burnup calculation module, construct a core diffusion and burnup calculation model, accurately solve the burnup equation, and update the nucleon density; S5. Update the macroscopic cross-section, use the component homogenized microscopic cross-section at the initial moment and the nucleon density updated by the component burnup calculation to assemble the component homogenized macroscopic cross-section of the next burnup step, and then perform the core calculation of the next burnup step until the end of the core life cycle.

2. The construction method of a fast fuel management model for a liquid metal fast reactor according to claim 1, characterized in that The specific implementation method of step S1 includes the following steps: S1.

1. Establish a two-dimensional single-component model through the CSG method. The two-dimensional single-component model applies a fully reflective boundary and is established according to the actual fuel assembly layout; S1.

2. Establish a super-component model through the CSG method. The super-component model applies a fully reflective boundary and is established by surrounding a non-fissile region component with a loop of fuel assemblies. The non-fissile region includes one of a bottom structure, an axial reflector, a gas chamber, a control component empty channel, a control component absorber, and a shielding layer; S1.

3. Establish a radial reflector model through the CSG method. The radial reflector model is a model with the fuel assembly in the middle and the radial reflector assembly on the outside, generating the homogenized group constants of the radial reflector region. The left and right sides of the radial reflector model adopt vacuum boundaries, and the upper and lower sides adopt reflective boundaries.

3. A method for constructing a fast fuel management model for a liquid metal fast reactor according to claim 1 or 2, characterized in that, The components described in step S1 belong to one of the lead-based traveling wave reactor, lead-cooled fast reactor, sodium-cooled fast reactor, and lead-bismuth fast reactor.

4. The construction method of a fast fuel management model for a liquid metal fast reactor according to claim 3, wherein Step S3 uses the homogenized macroscopic cross-sections of each component region to perform core nodal diffusion calculations, solve the three-dimensional full-core neutron diffusion equation, and obtain the core effective multiplication factor, power distribution, and neutron flux distribution results. S3.

1. The core nodal diffusion program is used to solve quadrilateral and hexagonal nodes, converting the solution of the three-dimensional neutron diffusion equation into the solution of three one-dimensional transverse integral equations. Among them, the nodal expansion method NEM is mainly used to solve the quadrilateral geometric nodes, and the equations are coupled through the transverse leakage term. S3.

2. For hexagonal geometric nodes, the NEM is used to solve the neutron diffusion equation axially. Radially, the hexagonal node is divided into six triangular nodes, and a local coordinate is set for each triangular node. The triangular polynomial expansion nodal TPEN method is used for solution.

5. The construction method of a fast fuel management model for a liquid metal fast reactor according to claim 4, wherein The specific implementation method of step S4 includes the following steps: S4.

1. In the initial state, based on the Monte Carlo program OpenMC, track the transmutation reactions in the burnup chain file to generate the microscopic parameter library required for independent burnup calculations. The transmutation reactions include fission, (n,γ), (n,2n), (n,3n), (n,4n), (n,p), and (n,α). S4.

2. Use the microscopic parameter library, burnup chain file, and power distribution results obtained from the core diffusion calculation at the current burnup step to perform component independent burnup calculations. OpenMC uses the Chebyshev rational approximation method to solve the matrix exponential, solve the Bateman equation, and update the nucleon density.

6. The construction method of a fast fuel management model for a liquid metal fast reactor according to claim 5, characterized in that, The specific process of updating the homogenized macroscopic cross-section at the burnup time point in step S5 is as follows: Based on the condition that the microscopic cross-section remains unchanged, assemble the new component homogenized macroscopic cross-section using the nucleon density of the component at the next burnup moment and the initial component homogenized microscopic cross-section, or assemble the new macroscopic cross-section using the nucleon density of the component at the next burnup moment and the homogenized microscopic cross-section. The formula for macroscopic cross-section assembly is: where x represents a specific reaction type, g is the coarse energy group number, i represents the nuclide, and Σ x,g,n represents the macroscopic cross section of the specific reaction in the homogenized region n and the g-th energy group, and N i,n represents the nucleon density of nuclide i in the homogenized region n; After obtaining the new component homogenized macroscopic cross-section, perform the core nodal diffusion calculation for the next burnup step.

7. A fast fuel management model for a liquid metal fast reactor, implemented relying on the construction method of a fast fuel management model for a liquid metal fast reactor according to one of claims 1-6, characterized in that, The described fast fuel management model for a liquid metal fast reactor assembles the component homogenized macroscopic cross-section for the next burnup step using the initial component homogenized microscopic cross-section and the nucleon density updated by component burnup calculations.

8. Application of a fast fuel management model for a liquid metal fast reactor, which is realized based on the method for constructing a fast fuel management model for a liquid metal fast reactor according to any one of claims 1-6, characterized in that For nuclear reactors, including lead-based traveling wave reactors, lead-cooled fast reactors, sodium-cooled fast reactors, and lead-bismuth fast reactors.

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