Rapid fuel management model for liquid metal fast reactor and construction method and application thereof

Abstract

The invention discloses a liquid metal fast reactor-oriented fast fuel management model as well as a construction method and application thereof, and belongs to the technical field of nuclear reactor physics. In order to realize rapid three-dimensional reactor core calculation, a two-dimensional single-component model, a super-component model and a radial reflecting layer model are established by constructing a solid geometry method; component calculation is carried out by using a Monte Carlo continuous energy point section neutron transport method; reactor core calculation: carrying out reactor core calculation by utilizing a nodal diffusion method, and solving a reactor core neutron diffusion equation; in combination with an independent burnup calculation module, constructing a reactor core diffusion and burnup calculation model, accurately solving a burnup equation, and updating the nuclear density; and updating the macroscopic cross section, calculating the updated nucleon density by using the component homogenized microscopic cross section and the component burnup at the initial moment, and assembling the component homogenized macroscopic cross section of the next burnup step to execute the reactor core calculation of the next burnup step until the end of the service life of the reactor core. According to the method, rapid reactor core burnup calculation is realized, and rapid fuel management is realized.

Description

Technical Field

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

[0002] Neutronic calculation and analysis is the core content of reactor core design and development. Usually, the Monte Carlo method or deterministic method is used to solve the neutron diffusion equation or transport equation. At present, the one-step method or the two-step method can be used to perform neutronic analysis on fast reactors. The one-step method directly solves the three-dimensional whole reactor, introduces fewer approximations, has accurate modeling, and high calculation accuracy. However, the cost of three-dimensional calculation of the core is huge. In fast reactor neutronic calculations (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 minority group constants of each component area, and then use the core program to perform three-dimensional core neutronic calculations.

[0003] Since medium-mass nuclides such as sodium and iron have strong elastic scattering resonance in the fast neutron region, the resonance effect in fast reactors is more complex than that in pressurized water reactors, and a more sophisticated energy group structure is required to analyze the energy self-screening effect. In addition, fast reactors have the characteristics of hard energy spectrum and no longer have the characteristics of "1 / E" spectrum. The two-step calculation model of pressurized water reactors cannot be directly used for fast reactor neutron analysis. Therefore, it is necessary to develop a two-step calculation model specifically for fast reactor neutron calculation analysis. Moreover, the current research on the two-step method of fast reactors based on Monte Carlo and core block methods at home and abroad mainly focuses on the calculation and analysis of initial steady-state conditions, and there is little research on microscopic fuel consumption calculation methods. Summary of the invention

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

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

[0006] A method for constructing a fast fuel management model for a liquid metal fast reactor comprises the following steps:

[0007] S1. Construct component models, establish two-dimensional single component models, super component models and radial reflection layer models by constructing solid geometry methods;

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

[0009] S3. Core calculation, using the nodal diffusion method to solve the core neutron diffusion equation;

[0010] 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;

[0011] 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.

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

[0013] S1.1. A two-dimensional single component model is established by using the CSG method. The two-dimensional single component model applies a total reflection boundary. The two-dimensional single component model is established for the actual fuel assembly arrangement.

[0014] S1.2. A superassembly model is established by using a CSG method. The superassembly model uses a total reflection boundary. The superassembly model is established by surrounding a non-breeding region assembly with a circle of fuel assemblies. The non-breeding region includes one of a bottom structure, an axial reflection layer, an air chamber, a control assembly empty channel, a control assembly absorber, and a shielding layer.

[0015] S1.3. A radial reflection layer model is established by the CSG method. The radial reflection layer model is a radial reflection layer model in which the fuel assembly is in the middle and the radial reflection layer assembly is on the outside. The homogenization group constants of the radial reflection layer area are generated. The radial reflection layer model uses vacuum boundaries on the left and right sides and reflection boundaries on the upper and lower sides.

[0016] Furthermore, the component in step S1 is 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. Using 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, the homogenization group constants of the fuel assembly area and each non-breeding area are generated through the assembly model constructed in step S1;

[0019] First, let the volume V of the homogenized region n be n The volume V of each calculation area m m The composition is expressed as:

[0020]

[0021] Among them, Σ is the summation operator. At the same time, the coarse energy group is composed of fine energy groups. In order to generate multi-group cross sections, the spatial region and energy group are merged. In the process of homogenizing and merging energy groups and regions, the reaction rate before and after the merging is guaranteed to be conserved, and then:

[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 the neutron reaction, e' represents the fine energy group number before the neutron reaction, ν represents the average number of neutrons produced per fission, Σ t,e,m is the average macroscopic total cross section of the e-th energy group in the calculation area 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 energy group in the spatially homogenized and grouped calculation region m, Σ s,e'→e,m For spatial homogenization, the average macroscopic scattering cross section from the e'th energy group to the eth energy group in the region m is calculated. 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 spatially homogenized and grouped homogenized region n, is the average macroscopic scattering cross section from the g'th coarse energy group to the gth coarse energy group in the homogenized region n after spatial homogenization and group merging, φ e,m The average neutron flux density of the e-th energy group in the region m is calculated 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. The average flux after homogenization is calculated from the flux before homogenization. The average cross section after homogenization is expressed as:

[0025]

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

[0027]

[0028] S2.3. For the super-component model, based on the influence of materials and neutron flux distribution, the homogenization calculation process is corrected by the super homogenization method SPH: the two-loop super-component model is used for calculation, and the SPH factor is iteratively calculated until convergence;

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

[0030]

[0031] Among them, r is a space vector, is the gradient operator, is the divergence operator, n is the homogenization 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, and χ g is the fraction of neutrons 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 the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic total cross section of the g-th energy group in the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic fission cross section of the g-th energy group in the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic scattering cross section from the g'th energy group to the gth energy group in the homogenized region n calculated by the Monte Carlo program OpenMC;

[0032] S2.3.2. Use the core block diffusion program to calculate the equivalent uniform two-circle assembly model to solve the multi-group neutron diffusion equation and obtain the average flux of the homogenized energy group g and block area n Then the SPH factor is calculated using the following formula:

[0033]

[0034] Where Σ is the summation operator, n is the homogenization region number, 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 block diffusion program. represents the normalized SPH factor of the homogenized region n and the g-th energy group, represents the average flux of the component homogenization region n and the g-th energy group calculated by the Monte Carlo program, represents the average flux of the component homogenization region n and the gth 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 the specific reaction type, n is the homogenization region number, g is the coarse 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 a specific reaction in homogenized region n and energy group g, The macroscopic cross section represents the specific reaction in 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, repeat steps S2.3.2 and S2.3.3 until the given convergence conditions are met, and obtain the final SPH factor and the SPH-corrected homogenized few-group constants.

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

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

[0041] S3.2. For hexagonal geometric nodes, NEM is used to solve the neutron diffusion equation in the axial direction, and the hexagonal node is divided into six triangular nodes in the radial direction. A local coordinate is set for each triangular node, and the triangular polynomial expansion node TPEN method is used to solve it.

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

[0043] 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);

[0044] 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.

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

[0046] Based on the condition that the microscopic cross section remains unchanged, the component nucleon density at the next burnup time and the component homogenized microscopic cross section at the initial time are used to assemble a new component homogenized macroscopic cross section, or the component nucleon density at the next burnup time and the homogenized microscopic cross section are used to assemble a new macroscopic cross section; the formula for macroscopic cross section assembly is:

[0047]

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

[0049] After obtaining the new component homogenized macroscopic cross section, the core segment diffusion calculation for the next burnup step is performed.

[0050] A fast fuel management model for liquid metal fast reactors is realized based on a method for constructing a fast fuel management model for liquid metal fast reactors. The fast fuel management model for liquid metal fast reactors utilizes the homogenized microscopic cross-section of components at the initial moment and the nuclear density updated by the component burnup calculation to assemble the homogenized macroscopic cross-section of components for the next burnup step.

[0051] An application of a fast fuel management model for liquid metal fast reactors 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] Beneficial effects of the present invention:

[0053] The invention discloses a method for constructing a fast fuel management model for a liquid metal fast reactor. The method uses the Monte Carlo method to solve the neutron transport equation to calculate the component homogenization group constant, uses a more refined continuous energy neutron cross section to analyze the energy self-screening effect, can accurately simulate the resonance interference phenomenon in the fast reactor, and generate a homogenization group constant with higher accuracy. Moreover, it is based on constructing solid geometry modeling and has strong geometric adaptability. The calculated component homogenization group constant is provided to the core block diffusion program to solve the neutron diffusion equation, realize fast three-dimensional core calculation, and obtain the results such as the core effective proliferation coefficient and power distribution.

[0054] The method for constructing a fast fuel management model for a liquid metal fast reactor described in the present invention uses a pre-generated microscopic parameter library for burnup calculation and the component power distribution obtained by core diffusion calculation at each burnup step to perform independent component burnup calculation, update nuclear density, realize fast core burnup calculation, and ultimately realize fast fuel management.

[0055] The method for constructing a fast fuel management model for a liquid metal fast reactor described in the present invention can assemble a component homogenized macroscopic cross-section for the next burnup step using the component homogenized microscopic cross-section at the initial moment and the nuclear density updated by the component burnup calculation, without the need to generate a large number of homogenized macroscopic cross-sections at the burnup moments in advance for interpolation calculation. BRIEF DESCRIPTION OF THE DRAWINGS

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

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

[0058] Figure 3 The two-dimensional single component model and super component model used for calculating the homogenization parameters of the Monte Carlo assembly in Example 1 of the present invention; wherein (a) is a two-dimensional single component model used for calculating the homogenization parameters of the fuel assembly, (b)-(h) are two-dimensional super component models used for calculating the homogenization parameters of the non-breeding region, (b) is an empty channel region super component model, (c) is an absorber region super component model, (d) is a shielding region super component model, (e) is a coolant region super component model, (f) is a bottom structure region super component model, (g) is an axial reflection layer region super component model, and (h) is a gas chamber region super component model;

[0059] Figure 4 It is a radial reflection layer model used for Monte Carlo component homogenization calculation in Example 1 of the present invention;

[0060] Figure 5is a component model used to calculate the super homogenization (SPH) factor in Example 1 of the present invention, wherein (a) is a non-uniform two-turn super component model, and (b) is an equivalent uniform two-turn super component model;

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

[0062] Figure 7 The deviation of the effective proliferation factor of the core obtained by assembling the macroscopic cross section by using the microscopic cross section that changes with the burnup in Example 1 of the present invention and performing the burnup calculation relative to the reference solution;

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

[0064] Fig. 9 The deviation of the effective proliferation factor of the core obtained by assembling the macroscopic cross-section using the microscopic cross-section that changes with the burnup and the nucleus density of the reference solution corresponding to the burnup step in Example 1 of the present invention relative to the reference solution. DETAILED DESCRIPTION

[0065] In order to make the purpose, technical solution and advantages of the present invention more clear, the present invention is further described in detail below in conjunction with the accompanying drawings and specific embodiments. It should be understood that the specific embodiments described herein are only used to explain the present invention and are not used to limit the present invention, that is, the specific embodiments described are only part of the embodiments of the present invention, rather than all of the specific embodiments. The components of the specific embodiments of the present invention described and shown in the drawings herein can be arranged and designed in various different configurations, and the present invention can also have other embodiments.

[0066] Therefore, the following detailed description of the specific embodiments of the present invention provided in the accompanying drawings is not intended to limit the scope of the invention claimed for protection, but merely represents the selected specific embodiments of the present invention. Based on the specific embodiments of the present invention, all other specific embodiments obtained by those skilled in the art without making creative work are within the scope of protection of the present invention.

[0067] In order to further understand the content, features and effects of the present invention, the following specific implementation methods are given as examples, and the attached Figure 1 -Attached Fig. 9 The detailed instructions are as follows:

[0068] Embodiment 1:

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

[0070] S1. Construct component models, establish two-dimensional single component models, super component models and radial reflection layer models by constructing solid geometry methods;

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

[0072] S1.1. A two-dimensional single component model is established by using the CSG method. The two-dimensional single component model applies a total reflection boundary. The two-dimensional single component model is established for the actual fuel assembly arrangement.

[0073] S1.2. A superassembly model is established by using a CSG method. The superassembly model uses a total reflection boundary. The superassembly model is established by surrounding a non-breeding region assembly with a circle of fuel assemblies. The non-breeding region includes one of a bottom structure, an axial reflection layer, an air chamber, a control assembly empty channel, a control assembly absorber, and a shielding layer.

[0074] S1.3. A radial reflection layer model is established by the CSG method. The radial reflection layer model is a radial reflection layer model in which the fuel assembly is in the middle and the radial reflection layer assembly is on the outside. The homogenization group constants of the radial reflection layer area are generated. The radial reflection layer model uses vacuum boundaries on the left and right sides and reflection boundaries on the upper and lower sides.

[0075] Furthermore, the component in step S1 is 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] Furthermore, the fuel assembly can not only provide neutrons for the calculation of homogenization group constants in the non-breeding region, 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 (standing wave reactor) oriented to 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 material switching scheme.

[0078] In order to deal with the complex resonance calculation problems in the fast neutron energy region, the continuous energy point cross-section data required for the Monte Carlo method to solve the neutron transport equation are given by the currently public evaluation nuclear database.

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

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

[0081] S2.1. Using 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, the homogenization group constants of the fuel assembly area and each non-breeding area are generated through the assembly model constructed in step S1;

[0082] First, let the volume V of the homogenized region n be n The volume V of each calculation area m m The composition is expressed as:

[0083]

[0084] Among them, Σ is the summation operator. At the same time, the coarse energy group is composed of fine energy groups. In order to generate multi-group cross sections, the spatial region and energy group are merged. In the process of homogenizing and merging energy groups and regions, the reaction rate before and after the merging is guaranteed to be conserved, and then:

[0085]

[0086] 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, Σ t,e,m is the average macroscopic total cross section of the e-th energy group in the calculation area 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 energy group in the spatially homogenized and grouped calculation region m, Σ s,e'→e,m For spatial homogenization, the average macroscopic scattering cross section from the e'th energy group to the eth energy group in the region m is calculated. 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 spatially homogenized and grouped homogenized region n, is the average macroscopic scattering cross section from the g'th coarse energy group to the gth coarse energy group in the homogenized region n after spatial homogenization and group merging, φ e,m The average neutron flux density of the e-th energy group in the region m is calculated 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. The average flux after homogenization is calculated from the flux before homogenization. The average cross section after homogenization is expressed as:

[0088]

[0089] When the spatial homogenization merged region is one region, this equivalence does not introduce approximation. However, for the problem of merging into two regions (super-component model), especially when the material difference is very large, such as calculating the super-component model of the absorber region ( Figure 3 (c)), due to the strong absorption region, the neutron flux distribution has a large difference, and the above homogenization calculation process has a certain approximation, which needs to be corrected by the SPH method. For the fission neutron energy spectrum, the actual homogenization calculation does not introduce approximation, and the reaction rate before and after homogenization is conserved, so there is no need for SPH correction.

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

[0091]

[0092] The key to SPH correction is the calculation of SPH factor. Due to the geometric modeling limitations of some core node diffusion programs, it is impossible to calculate equivalently. Figure 3 (b)-(h) show the supercomponent model, so the two-loop supercomponent model ( Figure 5 ) is used for calculation. SPH factors usually need to be iterated until convergence, such as Figure 6 As shown,

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

[0094] S2.3.1. Use the Monte Carlo neutron transport calculation program OpenMC to calculate the non-uniform two-circle component model, solve the neutron transport equation, and obtain the average neutron flux density and homogenized macroscopic cross-section results of the energy group g and the component homogenization area n. Fill the homogenized small group cross-section and fission neutron energy spectrum data into the diffusion program to construct the multi-group neutron diffusion equation:

[0095]

[0096] Among them, r is a space vector, is the gradient operator, is the divergence operator, n is the homogenization 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, and χ g is the fraction of neutrons 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 the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic total cross section of the g-th energy group in the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic fission cross section of the g-th energy group in the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic scattering cross section from the g'th energy group to the gth energy group in the homogenized region n calculated by the Monte Carlo program OpenMC;

[0097] S2.3.2. Use the core block diffusion program to calculate the equivalent uniform two-circle assembly model to solve the multi-group neutron diffusion equation and obtain the average flux of the homogenized energy group g and block area n Then the SPH factor is calculated using the following formula:

[0098]

[0099] Where Σ is the summation operator, n is the homogenization region number, g is the 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 node diffusion program. represents the normalized SPH factor of the homogenized region n and the g-th energy group, represents the average flux of the component homogenization region n and the g-th energy group calculated by the Monte Carlo program, represents the average flux of the component homogenization region n and the gth energy group calculated by the core diffusion program, N g is the normalization factor, V n represents the volume of the homogenized region n;

[0100] Use the calculated SPH factor to correct the macroscopic cross section:

[0101]

[0102] Among them, x represents the specific reaction type, n is the homogenization area 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 a specific reaction in homogenized region n and energy group g, The macroscopic cross section represents the specific reaction in 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 Correct the homogenized few-group constants of each energy group and each region, fill the corrected group constants into the diffusion program, repeat steps S2.3.2 and S2.3.3 until the given convergence conditions are met, and obtain the final SPH factor and the SPH-corrected homogenized few-group constants.

[0104] S3. Core calculation, using the nodal diffusion method to solve the core neutron diffusion equation;

[0105] Furthermore, step S3 uses the homogenized macroscopic cross-section of each component region to perform core block diffusion calculation, solve the three-dimensional full-reactor neutron diffusion equation, and obtain the core effective multiplication coefficient, power distribution, and neutron flux distribution results;

[0106] S3.1. The core node diffusion program is used to solve quadrilateral nodes and hexagonal nodes, converting the solution of the three-dimensional neutron diffusion equation into the solution of three one-dimensional transverse integral equations. The node 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 geometric nodes, NEM is used to solve the neutron diffusion equation in the axial direction, and the hexagonal node is divided into six triangular nodes in the radial direction. A local coordinate is set for each triangular node, and the triangular polynomial expansion node TPEN method is used to solve it.

[0108] Furthermore, the main idea of ​​the TPEN method is to use the nodal expansion method (NEM) in the axial direction to solve the neutron diffusion equation, divide the hexagonal node into six triangular nodes in the radial direction, set a local coordinate for each triangular node, and use the external normal direction of each hexagonal surface as the positive x direction of the local coordinate. The transversely integrated neutron diffusion equation in a hexagonal node area consists of six radial equations and one axial equation. The form of the neutron balance equation group 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 that the microscopic cross section remains unchanged, the component nucleon density at the next burnup time and the component homogenized microscopic cross section at the initial time are used to assemble a new component homogenized macroscopic cross section, or the component nucleon density at the next burnup time and the homogenized microscopic cross section are used to assemble a new macroscopic cross section; the formula for macroscopic cross section assembly is:

[0121]

[0122] Where x represents the specific reaction type, n is the homogenization region number, g is the coarse energy group number, i represents the nuclide, Σ x,g,n represents the macroscopic cross section of a specific reaction in homogenized region n and energy group g, N i,n represents the nucleon density of nuclide i in homogenized region n;

[0123] After obtaining the new component homogenized macroscopic cross section, the core segment diffusion calculation for the next burnup step is performed.

[0124] The homogenized macroscopic cross section is an important homogenization parameter required for reactor core neutronics calculations. Its value is determined by the microscopic cross section of each nuclide and the nucleon density of each nuclide. Since the microscopic cross section of the nuclide and the nucleon density usually change with the reactor operation time (burnup time), the core calculation needs to use the homogenized macroscopic cross section corresponding to the current burnup time point. In the macroscopic burnup scheme, the component homogenized macroscopic cross sections of a series of burnup time points are usually calculated in advance and then the homogenized macroscopic cross section of the required burnup time point is obtained by interpolation. In the microscopic burnup scheme, after each core calculation is completed, the core neutron flux distribution and power distribution calculation results at the current burnup time point are used to perform the burnup calculation, and the nucleon density of each nuclide at the next burnup time point in each burnup area of ​​the core is obtained. The new nucleon density and the homogenized microscopic cross section are used to assemble a new homogenized macroscopic cross section, and then the new homogenized macroscopic cross section can be used for the next core calculation.

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

[0126] Figure 7 The deviation of the effective proliferation factor of the core obtained by assembling the macroscopic cross section with the microscopic cross section that changes with the burnup in the embodiment of the present invention relative to the reference solution is shown below. The following is a brief error analysis of the calculation results of the embodiment of the present invention. First, the nucleon density of the main nuclides is analyzed. Figure 8 The deviation of the nucleon density of the main nuclides in the embodiment 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 proliferation factor of the core is analyzed. Fig. 9 The deviation of the effective proliferation factor of the core obtained by assembling the macroscopic cross section using the microscopic cross section that changes with the burnup and the nucleon density of the corresponding burnup step of the reference solution in the embodiment of the present invention relative to the reference solution is shown in the calculation results. When the nucleon density of each burnup step is the same, the deviation of the effective proliferation factor of the core calculated by the embodiment of the present invention relative to the reference solution is maintained at around -100pcm.

[0127] Embodiment 2:

[0128] A fast fuel management model for a liquid metal fast reactor is implemented based on the method for constructing 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 utilizes the homogenized microscopic cross-section of the components at the initial moment and the nuclear density updated by the component burnup calculation to assemble the homogenized macroscopic cross-section of the components for the next burnup step.

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

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

[0131] Furthermore, the specific process of calculating the homogenization group constants of each component area of ​​the lead-cooled traveling wave reactor is as follows: first, Monte Carlo programs such as OpenMC and Serpent are used to perform neutron transport calculations based on continuous energy point cross sections in the evaluation nuclear database, and the two-dimensional single component model and super component model are used to generate the homogenization group constants of the lead-based traveling wave reactor fuel area and non-proliferation area. Due to the non-negligible energy spectrum shift effect at the interface between the drive component and the radial reflector, the homogenization group constant of the radial reflector area is calculated separately using the two-dimensional radial reflector model. And the calculation results of the homogenization group constants of the strong absorption area are corrected using appropriate homogenization correction methods such as the super homogenization (SPH) method to ensure the conservation of neutron reaction rates in each homogenization area before and after homogenization.

[0132] Furthermore, the calculation process of core diffusion and burnup of the lead-cooled traveling wave reactor is as follows: First, the homogenized group constants of each component area calculated at the initial moment are used to perform the core node diffusion calculation at the initial moment of the core, and obtain the core effective proliferation coefficient, power distribution and other results. Then, the power distribution and neutron flux distribution results obtained by the core calculation are provided to the independent burnup calculation program to solve the Bateman equation to obtain the nucleon density of each component at the next burnup moment. Then, the homogenized microscopic cross section at the initial moment (or the microscopic cross section that changes with burnup obtained by performing component burnup calculation in advance) is assembled with the nucleon density of each component at the next burnup moment obtained by the independent burnup calculation to generate a homogenized macroscopic cross section. After obtaining the new component homogenized macroscopic cross section, the core node diffusion calculation of the next burnup step can be performed, thereby realizing the core burnup calculation and fuel management.

[0133] Embodiment 3:

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

[0135] Embodiment 4:

[0136] A fast fuel management model for a liquid metal fast reactor is implemented based on the method for constructing 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 utilizes the homogenized microscopic cross-section of the components at the initial moment and the nuclear density updated by the component burnup calculation to assemble the homogenized macroscopic cross-section of the components for the next burnup step.

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

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

[0139] Furthermore, the specific process of calculating the homogenization group constants of each component area of ​​the lead-cooled fast reactor is as follows: first, Monte Carlo programs such as OpenMC and Serpent are used to perform neutron transport calculations based on continuous energy point cross sections in the evaluation nuclear database, and the homogenization group constants of the lead-based traveling wave reactor fuel area and non-proliferation area are generated using a two-dimensional single component model and a super component model, respectively. Due to the non-negligible energy spectrum shift effect at the interface between the drive component and the radial reflector, the homogenization group constant of the radial reflector area is calculated using a two-dimensional radial reflector model alone. And the calculation results of the homogenization group constants of the strong absorption area are corrected using appropriate homogenization correction methods such as the super homogenization (SPH) method to ensure the conservation of the neutron reaction rate in each homogenization area before and after homogenization.

[0140] Furthermore, the calculation process of core diffusion and burnup of lead-cooled fast reactor is as follows: First, the homogenized group constants of each component region calculated at the initial moment are used to perform core node diffusion calculation at the initial moment of the core, and obtain the core effective proliferation coefficient, power distribution and other results. Then, the power distribution and neutron flux distribution results obtained by the core calculation are provided to the independent burnup calculation program to solve the Bateman equation to obtain the nucleon density of each component at the next burnup moment. Then, the homogenized microscopic cross section at the initial moment (or the microscopic cross section that changes with burnup obtained by performing component burnup calculation in advance) is assembled with the nucleon density of each component at the next burnup moment obtained by the independent burnup calculation to generate a homogenized macroscopic cross section. After obtaining the new component homogenized macroscopic cross section, the core node diffusion calculation of the next burnup step can be performed, thereby realizing the core burnup calculation and fuel management.

[0141] Embodiment 5:

[0142] A fast fuel management model for a liquid metal fast reactor is implemented based on the method for constructing 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 utilizes the homogenized microscopic cross-section of the components at the initial moment and the nuclear density updated by the component burnup calculation to assemble the homogenized macroscopic cross-section of the components for the next burnup step.

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

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

[0145] Furthermore, the specific process of calculating the homogenization group constants of each component region of the sodium-cooled fast reactor is as follows: first, Monte Carlo programs such as OpenMC and Serpent are used to perform neutron transport calculations based on continuous energy point cross sections in the evaluation nuclear database, and the homogenization group constants of the lead-based traveling wave reactor fuel region and non-proliferation region are generated using a two-dimensional single component model and a super component model, respectively. Due to the non-negligible energy spectrum shift effect at the interface between the drive component and the radial reflector, the homogenization group constant of the radial reflector region is calculated using a two-dimensional radial reflector model alone. And the calculation results of the homogenization group constants of the strong absorption region are corrected using appropriate homogenization correction methods such as the super homogenization (SPH) method to ensure the conservation of the neutron reaction rate in each homogenization region before and after homogenization.

[0146] Furthermore, the calculation process of the core diffusion and burnup of the sodium-cooled fast reactor is as follows: First, the homogenized group constants of each component region calculated at the initial moment are used to perform the core node diffusion calculation at the initial moment of the core, and the core effective proliferation coefficient, power distribution and other results are obtained. Then, the power distribution and neutron flux distribution results obtained by the core calculation are provided to the independent burnup calculation program to solve the Bateman equation to obtain the nucleon density of each component at the next burnup moment. Then, the homogenized microscopic cross section at the initial moment (or the microscopic cross section that changes with burnup obtained by performing component burnup calculation in advance) is assembled with the nucleon density of each component at the next burnup moment obtained by the independent burnup calculation to generate a homogenized macroscopic cross section. After obtaining the new component homogenized macroscopic cross section, the core node diffusion calculation of the next burnup step can be performed, thereby realizing the core burnup calculation and fuel management.

[0147] Embodiment 6:

[0148] A fast fuel management model for a liquid metal fast reactor is implemented based on the method for constructing 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 utilizes the homogenized microscopic cross-section of the components at the initial moment and the nuclear density updated by the component burnup calculation to assemble the homogenized macroscopic cross-section of the components for the next burnup step.

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

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

[0151] Furthermore, the specific process of calculating the homogenization group constants of each component area of ​​the lead-bismuth fast reactor is as follows: first, Monte Carlo programs such as OpenMC and Serpent are used to perform neutron transport calculations based on continuous energy point cross sections in the evaluation nuclear database, and the two-dimensional single component model and super component model are used to generate the homogenization group constants of the lead-based traveling wave reactor fuel area and non-proliferation area. Due to the non-negligible energy spectrum shift effect at the interface between the drive component and the radial reflector, the homogenization group constant of the radial reflector area is calculated separately using the two-dimensional radial reflector model. And the calculation results of the homogenization group constants of the strong absorption area are corrected by using appropriate homogenization correction methods such as the super homogenization (SPH) method to ensure the conservation of neutron reaction rates in each homogenization area before and after homogenization.

[0152] Furthermore, the calculation process of the core diffusion and burnup of the lead-bismuth fast reactor is as follows: First, the homogenized group constants of each component region calculated at the initial moment are used to perform the core node diffusion calculation at the initial moment of the core, and the core effective proliferation coefficient, power distribution and other results are obtained. Then, the power distribution and neutron flux distribution results obtained by the core calculation are provided to the independent burnup calculation program to solve the Bateman equation to obtain the nucleon density of each component at the next burnup moment. Then, the homogenized microscopic cross section at the initial moment (or the microscopic cross section that changes with burnup obtained by performing component burnup calculation in advance) is assembled with the nucleon density of each component at the next burnup moment obtained by the independent burnup calculation to generate a homogenized macroscopic cross section. After obtaining the new component homogenized macroscopic cross section, the core node diffusion calculation of 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 used only to distinguish one entity or operation from another entity or operation, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Moreover, the terms "include", "comprise" or any other variants thereof are intended to cover non-exclusive inclusion, so that a process, method, article or device including a series of elements includes not only those elements, but also other elements not explicitly listed, or also includes elements inherent to such process, method, article or device. In the absence of further restrictions, the elements defined by the sentence "comprise a ..." do not exclude the existence of other identical elements in the process, method, article or device including the elements.

[0154] Although the present application has been described above with reference to specific embodiments, various modifications may be made thereto and parts thereof may 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 may be used in combination with each other in any manner, and the fact that these combinations are not exhaustively described in this specification is only for the sake of omitting space and saving 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: The steps include: S1. Construct component models, establish two-dimensional single component models, super component models and radial reflection layer models by constructing solid geometry methods; S2. Component calculation, using the Monte Carlo continuous energy point cross-section neutron transport method for component calculation; S3. Core calculation, using the nodal diffusion method to perform core calculation and solve the core neutron diffusion equation; 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; 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.

2. The method for constructing 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. A two-dimensional single component model is established by using the CSG method. The two-dimensional single component model applies a total reflection boundary. The two-dimensional single component model is established for the actual fuel assembly arrangement. S1.

2. A superassembly model is established by using a CSG method. The superassembly model uses a total reflection boundary. The superassembly model is established by surrounding a non-breeding region assembly with a circle of fuel assemblies. The non-breeding region includes one of a bottom structure, an axial reflection layer, an air chamber, a control assembly empty channel, a control assembly absorber, and a shielding layer. S1.

3. A radial reflection layer model is established by the CSG method. The radial reflection layer model is a radial reflection layer model in which the fuel assembly is in the middle and the radial reflection layer assembly is on the outside. The homogenization group constants of the radial reflection layer area are generated. The radial reflection layer model uses vacuum boundaries on the left and right sides and reflection boundaries on the upper and lower sides.

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 component in step S1 is one of a lead-based traveling wave reactor, a lead-cooled fast reactor, a sodium-cooled fast reactor and a lead-bismuth fast reactor.

4. The method for constructing a fast fuel management model for a liquid metal fast reactor according to claim 3, characterized in that: The specific implementation method of step S2 includes the following steps: S2.

1. Using 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, the homogenization group constants of the fuel assembly area and each non-breeding area are generated through the assembly model constructed in step S1; First, let the volume V of the homogenized region n be n The volume V of each calculation area m m The composition is expressed as: Among them, Σ is the summation operator. At the same time, the coarse energy group is composed of fine energy groups. In order to generate multi-group cross sections, the spatial region and energy group are merged. In the process of homogenizing and merging energy groups and regions, the reaction rate before and after the merging is guaranteed to be conserved, and then: 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, Σ t,e,m is the average macroscopic total cross section of the e-th energy group in the calculation area 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 energy group in the spatially homogenized and grouped calculation region m, Σ s,e'→e,m For spatial homogenization, the average macroscopic scattering cross section from the e'th energy group to the eth energy group in the region m is calculated. 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 spatially homogenized and grouped homogenized region n, is the average macroscopic scattering cross section from the g'th coarse energy group to the gth coarse energy group in the homogenized region n after spatial homogenization and group merging, φ e,m The average neutron flux density of the e-th energy group in the region m is calculated 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. The average flux after homogenization is calculated from the flux before homogenization. The average cross section after homogenization is expressed as: Based on the conservation of reaction rate before and after homogenization, the calculation formula of fission neutron spectrum 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 by the super homogenization method SPH: the two-loop super-component model is used for calculation, and the SPH factor is iteratively calculated until convergence; S2.3.

1. Use the Monte Carlo neutron transport calculation program OpenMC to calculate the non-uniform two-circle component model, solve the neutron transport equation, and obtain the average neutron flux density and homogenized macroscopic cross-section results of the energy group g and the component homogenization area n. Fill the homogenized small group cross-section and fission neutron energy spectrum data into the diffusion program to construct the multi-group neutron diffusion equation: Among them, r is a space vector, is the gradient operator, is the divergence operator, n is the homogenization 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, and χ g is the fraction of neutrons 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 the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic total cross section of the g-th energy group in the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic fission cross section of the g-th energy group in the homogenized region n calculated by the Monte Carlo program OpenMC, is the average macroscopic scattering cross section from the g'th energy group to the gth energy group in the homogenized region n calculated by the Monte Carlo program OpenMC; S2.3.

2. Use the core block diffusion program to calculate the equivalent uniform two-circle assembly model to solve the multi-group neutron diffusion equation and obtain the average flux of the homogenized energy group g and block area n Then the SPH factor is calculated using the following formula: Where Σ is the summation operator, n is the homogenization region number, 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 block diffusion program. represents the normalized SPH factor of the homogenized region n and the g-th energy group, represents the average flux of the component homogenization region n and the g-th energy group calculated by the Monte Carlo program, represents the average flux of the component homogenization region n and the gth 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 the specific reaction type, n is the homogenization region number, g is the coarse 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 a specific reaction in homogenized region n and energy group g, The macroscopic cross section represents the specific reaction in the homogenized region n and the g-th energy group after SPH correction; 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, repeat steps S2.3.2 and S2.3.3 until the given convergence conditions are met, and obtain the final SPH factor and the SPH-corrected homogenized few-group constants.

5. The method for constructing a fast fuel management model for a liquid metal fast reactor according to claim 4, characterized in that: Step S3 uses the homogenized macroscopic cross-section of each component area to perform core block diffusion calculation, solve the three-dimensional full-reactor neutron diffusion equation, and obtain the core effective multiplication coefficient, power distribution and neutron flux distribution results; S3.

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

2. For hexagonal geometric nodes, NEM is used to solve the neutron diffusion equation in the axial direction, and the hexagonal node is divided into six triangular nodes in the radial direction. A local coordinate is set for each triangular node, and the triangular polynomial expansion node TPEN method is used to solve it.

6. The method for constructing a fast fuel management model for a liquid metal fast reactor according to claim 5, characterized in that: The specific implementation method of step S4 includes the following steps: 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); 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.

7. The method for constructing a fast fuel management model for a liquid metal fast reactor according to claim 6, characterized in that: The specific process of updating the homogenized macroscopic cross section with the burnup time point in step S5 is: Based on the condition that the microscopic cross section remains unchanged, the component nucleon density at the next burnup time and the component homogenized microscopic cross section at the initial time are used to assemble a new component homogenized macroscopic cross section, or the component nucleon density at the next burnup time and the homogenized microscopic cross section are used to assemble a new macroscopic cross section; the formula for macroscopic cross section assembly is: Where x represents the specific reaction type, n is the homogenization region number, g is the coarse energy group number, i represents the nuclide, Σ x,g,n represents the macroscopic cross section of a specific reaction in homogenized region n and energy group g, N i,n represents the nucleon density of nuclide i in homogenized region n; After obtaining the new component homogenized macroscopic cross section, the core segment diffusion calculation for the next burnup step is performed.

8. A fast fuel management model for a liquid metal fast reactor, implemented by the method for constructing a fast fuel management model for a liquid metal fast reactor according to any one of claims 1 to 7, characterized in that: The fast fuel management model uses the component homogenized microscopic cross section at the initial moment and the nucleus density updated by the component burnup calculation to assemble the component homogenized macroscopic cross section of the next burnup step.

9. An application of a fast fuel management model for a liquid metal fast reactor, implemented by relying on a method for constructing a fast fuel management model for a liquid metal fast reactor according to any one of claims 1 to 7, characterized in that: Used in 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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